Design and development of novel MRI compatible zirconium- ruthenium alloys with ultralow magnetic susceptibility.

PubMed

Li, H F; Zhou, F Y; Li, L; Zheng, Y F

2016-04-19

In the present study, novel MRI compatible zirconium-ruthenium alloys with ultralow magnetic susceptibility were developed for biomedical and therapeutic devices under MRI diagnostics environments. The results demonstrated that alloying with ruthenium into pure zirconium would significantly increase the strength and hardness properties. The corrosion resistance of zirconium-ruthenium alloys increased significantly. High cell viability could be found and healthy cell morphology observed when culturing MG 63 osteoblast-like cells and L-929 fibroblast cells with zirconium-ruthenium alloys, whereas the hemolysis rates of zirconium-ruthenium alloys are <1%, much lower than 5%, the safe value for biomaterials according to ISO 10993-4 standard. Compared with conventional biomedical 316L stainless steel, Co-Cr alloys and Ti-based alloys, the magnetic susceptibilities of the zirconium-ruthenium alloys (1.25 × 10(-6) cm(3)·g(-1)-1.29 × 10(-6) cm(3)·g(-1) for zirconium-ruthenium alloys) are ultralow, about one-third that of Ti-based alloys (Ti-6Al-4V, ~3.5 × 10(-6) cm(3)·g(-1), CP Ti and Ti-6Al-7Nb, ~3.0 × 10(-6) cm(3)·g(-1)), and one-sixth that of Co-Cr alloys (Co-Cr-Mo, ~7.7 × 10(-6) cm(3)·g(-1)). Among the Zr-Ru alloy series, Zr-1Ru demonstrates enhanced mechanical properties, excellent corrosion resistance and cell viability with lowest magnetic susceptibility, and thus is the optimal Zr-Ru alloy system as therapeutic devices under MRI diagnostics environments.

Design and development of novel MRI compatible zirconium- ruthenium alloys with ultralow magnetic susceptibility

PubMed Central

Li, H.F.; Zhou, F.Y.; Li, L.; Zheng, Y.F.

2016-01-01

In the present study, novel MRI compatible zirconium-ruthenium alloys with ultralow magnetic susceptibility were developed for biomedical and therapeutic devices under MRI diagnostics environments. The results demonstrated that alloying with ruthenium into pure zirconium would significantly increase the strength and hardness properties. The corrosion resistance of zirconium-ruthenium alloys increased significantly. High cell viability could be found and healthy cell morphology observed when culturing MG 63 osteoblast-like cells and L-929 fibroblast cells with zirconium-ruthenium alloys, whereas the hemolysis rates of zirconium-ruthenium alloys are <1%, much lower than 5%, the safe value for biomaterials according to ISO 10993-4 standard. Compared with conventional biomedical 316L stainless steel, Co–Cr alloys and Ti-based alloys, the magnetic susceptibilities of the zirconium-ruthenium alloys (1.25 × 10−6 cm3·g−1–1.29 × 10−6 cm3·g−1 for zirconium-ruthenium alloys) are ultralow, about one-third that of Ti-based alloys (Ti–6Al–4V, ~3.5 × 10−6 cm3·g−1, CP Ti and Ti–6Al–7Nb, ~3.0 × 10−6 cm3·g−1), and one-sixth that of Co–Cr alloys (Co–Cr–Mo, ~7.7 × 10−6 cm3·g−1). Among the Zr–Ru alloy series, Zr–1Ru demonstrates enhanced mechanical properties, excellent corrosion resistance and cell viability with lowest magnetic susceptibility, and thus is the optimal Zr–Ru alloy system as therapeutic devices under MRI diagnostics environments. PMID:27090955

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.

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)

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.

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.

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.

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.

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

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.

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.

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.

Method of making crack-free zirconium hydride

DOEpatents

Sullivan, Richard W.

1980-01-01

Crack-free hydrides of zirconium and zirconium-uranium alloys are produced by alloying the zirconium or zirconium-uranium alloy with beryllium, or nickel, or beryllium and scandium, or nickel and scandium, or beryllium and nickel, or beryllium, nickel and scandium and thereafter hydriding.

Comparison of palladium and zirconium treated graphite tubes for in-atomizer trapping of hydrogen selenide in hydride generation electrothermal atomization atomic absorption spectrometry

NASA Astrophysics Data System (ADS)

Laborda, Francisco; Medrano, Jesús; Cortés, José I.; Mir, José M.; Castillo, Juan R.

1999-02-01

Zirconium treated graphite tubes were investigated and compared with non-treated and palladium coated ones for in situ trapping of selenium hydride generated in a flow injection system. Selenium was effectively trapped on zirconium treated tubes at trapping temperatures of 300-600°C, similar to those observed for palladium, whereas trapping temperatures higher than 600°C had to be used with non-treated tubes. Zirconium treated tubes used in this work showed good stability up to 300 trapping/atomization cycles, with precision better than 5%, characteristic masses of 42 (peak height) and 133 pg (peak area) of selenium were obtained. Sensitivity of zirconium and palladium treatments were similar, but zirconium offered the advantage of a single application per tube. Detection limits were 0.11 (peak height) and 0.23 ng (peak area) for a 1 ml sample volume.

White Paper Summary of 2nd ASTM International Workshop on Hydrides in Zirconium Alloy Cladding

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

Sindelar, R.; Louthan, M.; PNNL, B.

2015-05-29

This white paper recommends that ASTM International develop standards to address the potential impact of hydrides on the long term performance of irradiated zirconium alloys. The need for such standards was apparent during the 2nd ASTM International Workshop on Hydrides in Zirconium Alloy Cladding and Assembly Components, sponsored by ASTM International Committee C26.13 and held on June 10-12, 2014, in Jackson, Wyoming. The potentially adverse impacts of hydrogen and hydrides on the long term performance of irradiated zirconium-alloy cladding on used fuel were shown to depend on multiple factors such as alloy chemistry and processing, irradiation and post irradiation history,more » residual and applied stresses and stress states, and the service environment. These factors determine the hydrogen content and hydride morphology in the alloy, which, in turn, influence the response of the alloy to the thermo-mechanical conditions imposed (and anticipated) during storage, transport and disposal of used nuclear fuel. Workshop presentations and discussions showed that although hydrogen/hydride induced degradation of zirconium alloys may be of concern, the potential for occurrence and the extent of anticipated degradation vary throughout the nuclear industry because of the variations in hydrogen content, hydride morphology, alloy chemistry and irradiation conditions. The tools and techniques used to characterize hydrides and hydride morphologies and their impacts on material performance also vary. Such variations make site-to-site comparisons of test results and observations difficult. There is no consensus that a single material or system characteristic (e.g., reactor type, burnup, hydrogen content, end-of life stress, alloy type, drying temperature, etc.) is an effective predictor of material response during long term storage or of performance after long term storage. Multi-variable correlations made for one alloy may not represent the behavior of another alloy exposed

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.

Production of FR Tubing from Advanced ODS Alloys

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

Maloy, Stuart Andrew; Lavender, Curt; Omberg, Ron

2016-10-25

Significant research is underway to develop LWR nuclear fuels with improved accident tolerance. One of the leading candidate materials for cladding are the FeCrAl alloys. New alloys produced at ORNL called Gen I and Gen II FeCrAl alloys possess excellent oxidation resistance in steam up to 1400°C and in parallel methods are being developed to produce tubing from these alloys. Century tubing continues to produce excellent tubing from FeCrAl alloys. This memo reports receipt of ~21 feet of Gen I FeCrAl alloy tubing. This tubing will be used for future tests including burst testing, mechanical testing and irradiation testing.

Production of intergranular attack of alloy 600, alloy 690, and alloy 800 tubing in tubesheet crevices: Topical report

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

Scott, D.B.; Glaves, C.L.,

1987-07-01

Three model boilers, manufactured to simulate full-size tube sheet crevices, were tested with various secondary side environments. The first was faulted with organics representative of the decomposition of humic acid. The second was faulted with sodium carbonate and sodium hydroxide, while the third was faulted with sodium sulfate and sodium hydroxide. Each model contained seven tubes, which included Alloy 600 in the mill-annealed (MA) and thermally-treated (TT) conditions and Alloy 690 in the thermally-treated condition. Two models contained Alloy 800 tubes in the mill-annealed condition and one had Alloy 800 in the mill-annealed/cold-worked/glass-bead-peened condition. Two different sizes of tubesheet crevicesmore » were used in all model boilers. In the organics-faulted boiler, tubes of Alloy 600MA, Alloy 690TT and Alloy 800MA experienced no significant intergranular attack (IGA); however, the Alloy 600TT had intergranular attack (IGA) three to four grains deep. The carbonate-caustic faulted boiler experienced throughwall stress corrosion cracking (SCC) in all tubes of Alloy 600 MA and Alloy 800 MA. Eddy current indications were present in Alloy 690TT, Alloy 600TT and Alloy 800 in the mill-annealed/cold worked/glass-bead-peened condition. Metallographic examination of tubes from the third model boiler, faulted with sodium sulfate and caustic, revealed IGA in the mill-annealed Alloy 600 tubes. The IGA was more extensive in an Alloy 600 tube annealed at 1700/sup 0/F than an Alloy 600 tube annealed at 1875/sup 0/F.« less

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.

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

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.

Alloy 690 for steam generator tubing applications

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

Gold, R.E.; Harrod, D.L.; Aspden, R.G.

1990-10-01

This report has been prepared to provide background information for Ni-Cr-Fe Alloy 690 which is currently the material of choice for steam generator heat transfer tubing applications. Activities directed toward the qualification of Alloy 690 for these applications are summarized; this includes efforts which focused on optimization of materials procurement specifications. Emphasis is placed on research accomplished primarily in the four year period from June 1985, the time of the first EPRI Workshop on Alloy 690 was held. The topic is treated in a broad sense, and includes review of the physical metallurgy of the alloy, tube manufacturing processes, themore » properties of commercial production tubing, and the corrosion behavior of Alloy 690 in environments appropriate to steam generator service. 12 refs., 7 figs., 8 tabs.« less

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)

Crystal plasticity simulation of Zirconium tube rolling using multi-grain representative volume element

NASA Astrophysics Data System (ADS)

Isaenkova, Margarita; Perlovich, Yuriy; Zhuk, Dmitry; Krymskaya, Olga

2017-10-01

The rolling of Zirconium tube is studied by means of the crystal plasticity viscoplastic self-consistent (VPSC) constitutive modeling. This modeling performed by a dislocation-based constitutive model and a spectral solver using open-source simulation of DAMASK kit. The multi-grain representative volume elements with periodic boundary conditions are used to predict the texture evolution and distributions of strain and stresses. Two models for randomly textured and partially rolled material are deformed to 30% reduction in tube wall thickness and 7% reduction in tube diameter. The resulting shapes of the models are shown and distributions of strain are plotted. Also, evolution of grain's shape during deformation is shown.

Spring/dimple instrument tube restraint

DOEpatents

DeMario, Edmund E.; Lawson, Charles N.

1993-01-01

A nuclear fuel assembly for a pressurized water nuclear reactor has a spring and dimple structure formed in a non-radioactive insert tube placed in the top of a sensor receiving instrumentation tube thimble disposed in the fuel assembly and attached at a top nozzle, a bottom nozzle, and intermediate grids. The instrumentation tube thimble is open at the top, where the sensor or its connection extends through the cooling water for coupling to a sensor signal processor. The spring and dimple insert tube is mounted within the instrumentation tube thimble and extends downwardly adjacent the top. The springs and dimples restrain the sensor and its connections against lateral displacement causing impact with the instrumentation tube thimble due to the strong axial flow of cooling water. The instrumentation tube has a stainless steel outer sleeve and a zirconium alloy inner sleeve below the insert tube adjacent the top. The insert tube is relatively non-radioactivated inconel alloy. The opposed springs and dimples are formed on diametrically opposite inner walls of the insert tube, the springs being formed as spaced axial cuts in the insert tube, with a web of the insert tube between the cuts bowed radially inwardly for forming the spring, and the dimples being formed as radially inward protrusions opposed to the springs.

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

Spring/dimple instrument tube restraint

DOEpatents

DeMario, E.E.; Lawson, C.N.

1993-11-23

A nuclear fuel assembly for a pressurized water nuclear reactor has a spring and dimple structure formed in a non-radioactive insert tube placed in the top of a sensor receiving instrumentation tube thimble disposed in the fuel assembly and attached at a top nozzle, a bottom nozzle, and intermediate grids. The instrumentation tube thimble is open at the top, where the sensor or its connection extends through the cooling water for coupling to a sensor signal processor. The spring and dimple insert tube is mounted within the instrumentation tube thimble and extends downwardly adjacent the top. The springs and dimples restrain the sensor and its connections against lateral displacement causing impact with the instrumentation tube thimble due to the strong axial flow of cooling water. The instrumentation tube has a stainless steel outer sleeve and a zirconium alloy inner sleeve below the insert tube adjacent the top. The insert tube is relatively non-radioactivated inconel alloy. The opposed springs and dimples are formed on diametrically opposite inner walls of the insert tube, the springs being formed as spaced axial cuts in the insert tube, with a web of the insert tube between the cuts bowed radially inwardly for forming the spring, and the dimples being formed as radially inward protrusions opposed to the springs. 7 figures.

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.

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.

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

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.

Effects of the Fabrication Process and Thermal Cycling on the Oxidation of Zirconium-Niobium Pressure Tubes

NASA Astrophysics Data System (ADS)

Nam, Cheol

2009-12-01

Pressure tubes made of Zr-2.5%Nb alloy are used to contain fuels and coolant in CANDU nuclear power reactors The pressure tube oxidizes during reactor operation and hydrogen ingress through the oxide grown on the tube limits its lifetime. Little attention was paid to the intermediate tube manufacturing processes in enhancing the oxidation resistance. In addition, the oxide grown on the tube experiences various thermal cycles depending on the reactor shutdown and startup cycles. To address these two aspects and to better understand the oxidation process of the Zr-2.5Nb tube, research was conducted in two parts: (i) effects of tube fabrication on oxidation behavior, and (ii) thermal cycling behaviors of oxides grown on a pressure tube. In the first part, the optimum manufacturing process was pursued to improve the corrosion resistance of Zr-2.5Nb tubes. Experimental micro-tubes were fabricated with various manufacturing routes in the stages of billet preparation, hot extrusion and cold drawing. These were oxidized in air at 400°C and 500°C, and in an autoclave at 360°C lithiated water. Microstructure and texture of the tubes and oxides were characterized with X-ray diffraction, scanning electron microscope and optical microscope. Special emphasis was given to examinations of the metal/oxide interface structures. A correlation between the manufacturing process and oxidation resistance was investigated in terms of tube microstructure and the metal/oxide interface structure. As a result, it was consistently observed that uniform interface structures were formed on the tubes which had a fine distribution of secondary phases. These microstructures were found to be beneficial in enhancing the oxidation resistance as opposed to the tubes that had coarse and continuous beta-Zr phases. Based on these observations, a schematic model of the oxidation process was proposed with respect to the oxidation resistance under oxidizing temperatures of 360°C, 400°C and 500°C. In

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.

Torsional Stability of Aluminum Alloy Seamless Tubing

NASA Technical Reports Server (NTRS)

Moore, R L; Paul, D A

1939-01-01

Torsion tests were made on 51ST aluminum-alloy seamless tubes having diameter-to-thickness ratios of from 77 to 139 and length-to-diameter ratios of from 1 to 60. The torsional strengths developed in the tubes which failed elastically (all tubes having lengths greater than 2 to 6 times the diameter) were in most cases within 10 percent of the value indicated by the theories of Donnel, Timoshenko, and Sturm, assuming a condition of simply supported ends.

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.

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.

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.

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.

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.

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.

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.

Relationship between microstructure and mechanical properties of Alloy 690 tubes for steam generators

NASA Astrophysics Data System (ADS)

Diano, P.; Muggeo, A.; Van Duysen, J. C.; Guttmann, M.

1989-12-01

Alloy 690 is used to replace Alloy 600 for the fabrication of tubes for steam generators of french pressurized water nuclear reactors. In order to reduce the dispersion in tensile properties observed for the first Alloy 690 industrial tubes, and which had already been noticed for Alloy 600, a joint research programme has been carried out by Electricité de France (Département Etude des Matériaux) and Valinox Montbard. The dispersion in the tensile properties of the first industrial Alloy 690 tubes for PWR steam generators arises from two main factors: - a grain size dispersion which is in particular controlled by the carbon content and by the final heat treatment, - differences in the degree of strain hardening induced by the straightening process with rollers. An improvement of the fabrication processes which have an influence on these two factors has allowed to reduce considerably the dispersion of the tensile properties of the more recent series of tubes.

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.

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.

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.

Evaluation of candidate stirling engine heater tube alloys at 820 deg and 860 deg C

NASA Technical Reports Server (NTRS)

Misencik, J. A.

1982-01-01

Seven commercial alloys were evaluated in Stirling simulator materials rigs. Five iron base alloys (N-155, A-286, Incoloy 800, 19-9DL, and 316 stainless steel), one nickel base alloy (Inconel 718), and one cobalt base alloy (HS-188) were tested in the form of thin wall tubing in a diesel fuel fired test rig. Tubes filled with hydrogen or helium at gas pressure of 21.6 MPa and temperatures of 820 and 860 C were endurance tested for 1000 and 535 hours, respectively. Results showed that under these conditions hydrogen permeated rapidly through the tube walls, thus requiring refilling during each five hour cycle. Helium was readily contained, exhibiting no measurable loss by permeation. Helium filled tubes tested at 860 C all exhibited creep-rupture failures within the 535 hour endurance test. Subsequent tensile test evaluation after removal from the rig indicated reduced room temperature ductility for some hydrogen-filled tubes compared to helium-filled tubes, suggesting possible hydrogen embrittlement in these alloys.

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.

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.

Alloying effect of copper concentration on the localized corrosion of aluminum alloy for heat exchanger tube

NASA Astrophysics Data System (ADS)

Hong, Min-Sung; Park, In-Jun; Kim, Jung-Gu

2017-07-01

This study examined the alloying effect of Cu content on the localized corrosion properties of Al alloy in synthetic acid rain containing 200 ppm of Cl- ion. In aluminum alloy tubes, a small amount of Cu is contained as the additive to improve the mechanical strength or as the impurity. The Cu-containing intermetallic compound, Al2Cu can cause galvanic corrosion because it has more noble potential than Al matrix. Therefore aluminum tube could be penetrated by localized corrosion attack. The results were obtained from electrochemical test, scanning electron microscopy, and time of flight secondary ion mass spectrometry (ToF-SIMS) mapping. Severe localized corrosion was occurred on the Al-0.03 wt% Cu alloy. The negative effect of Cu on the pitting corrosion was attributed to the presence of the Al2Cu precipitates.

Evaluation of candidate Stirling engine heater tube alloys for 1000 hours at 760 C

NASA Technical Reports Server (NTRS)

Misencik, J. A.

1980-01-01

Six tubing alloys were endurance tested in a diesel fired, Stirling engine simulator materials test rig for 1000 hours of 760 C while pressurized at 17 to 21 MPa with either hydrogen or helium. The alloys tested were N 155, A 286, Incoloy 800, 19 9DL, Nitronic 40 and 316 stainless steel. The alloys were in the form of thin wall tubing. Hydrogen permeated rapidly through the tube walls of all six alloys when they were heated to 760 C. Helium was readily contained. Creep rupture failures occurred in four of the six alloys pressurized with hydrogen. Only two alloys survived the 1000 hour endurance test with no failures. Simultaneous exposure to either hydrogen or helium and the combustion environment did not seriously degrade the tensile strength of the six alloys in room temperature or 760 C tests after exposure. Decreases in room temperature ductility were observed and are attributed to aging rather than to hydrogen embrittlement in three of the alloys. However, there may be a hydrogen embrittlement effect in the N 155, 19 9DL, and Nitronic 40 alloys.

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

On the influence of manufacturing practices on the SCC behavior of Alloy 690 steam generator tubing

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

Doherty, P.E.; Doyle, D.M.; Sarver, J.M.

1996-12-31

Thermally treated (TT) Alloy 690 is the tubing materials of choice for replacement steam generators (RSGs) throughout the world. It is manufactured using a variety of processing methods with regards to melt practice and thermomechanical forming. Studies assessing the IGSCC resistance of Alloy 690 TT SG tubing have identified a variability in the corrosion performance of nominally identical alloys. While tubing of comparable bulk chemistry may exhibit variations in microchemistry as a result of different melt practice, the correlation between melt practice and SCC resistance is difficult to assess due to other contributing factors. The other contributing factors are identifiedmore » in this investigation as microstructural features whose generation is dependent on features of particular strain-anneal forming methods by which SG tubes are fabricated. In this study the microstructural characteristics which appear to affect inservice corrosion performance of Alloy 690 TT SG tubes were evaluated. The studies included extensive microstructural examinations in addition to CERT tests performed on actual Alloy 690 TT nuclear SG tubing. The CERT test results indicate that Alloy 690 TT tubing processed at higher mill anneal temperatures display the highest degree of stress corrosion cracking (SCC) resistance. This observation is discussed with reference to carbide distributions, textural aspects and grain boundary orientation character.« less

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.

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 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.

Fabrication of Titanium-Niobium-Zirconium-Tantalium Alloy (TNZT) Bioimplant Components with Controllable Porosity by Spark Plasma Sintering

PubMed Central

Rechtin, Jack; Torresani, Elisa; Ivanov, Eugene; Olevsky, Eugene

2018-01-01

Spark Plasma Sintering (SPS) is used to fabricate Titanium-Niobium-Zirconium-Tantalum alloy (TNZT) powder—based bioimplant components with controllable porosity. The developed densification maps show the effects of final SPS temperature, pressure, holding time, and initial particle size on final sample relative density. Correlations between the final sample density and mechanical properties of the fabricated TNZT components are also investigated and microstructural analysis of the processed material is conducted. A densification model is proposed and used to calculate the TNZT alloy creep activation energy. The obtained experimental data can be utilized for the optimized fabrication of TNZT components with specific microstructural and mechanical properties suitable for biomedical applications. PMID:29364165

The temperature dependence of the tensile properties of thermally treated Alloy 690 tubing

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

Harrod, D.L.; Gold, R.E.; Larsson, B.

1992-12-31

Tensile tests were run in air on full tube cross-sections of 22.23 mm OD by 1.27 mm wall thickness Alloy 690 steam generator production tubes from ten (10) heats of material at eight (8) temperatures between room temperature and 760{degrees}C. The tubing was manufactured to specification requirements consistent with the EPRI guidelines for Alloy 690 tubing. The room temperature stress-strain curves are described quite well by the Voce equation. Ductile fracture by dimpled rupture was observed at all test temperatures. The elevated temperature tensile properties are compared with design data given in the ASME Code.

Method of manufacturing corrosion resistant tubing from welded stock of titanium or titanium base alloy

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

Meredith, S.E.; Benjamin, J.F.

1993-07-13

A method is described of manufacturing corrosion resistant tubing from seam welded stock of a titanium or titanium based alloy, comprising: cold pilgering a seam welded tube hollow of titanium or titanium based alloy in a single pass to a final sized tubing, the tube hollow comprising a strip which has been bent and welded along opposed edges thereof to form the tube hollow, the tube hollow optionally being heat treated prior to the cold pilgering step provided the tube hollow is not heated to a temperature which would transform the titanium or titanium alloy into the beta phase, themore » cold pilgering effecting a reduction in cross sectional area of the tube hollow of at least 50% and a reduction of wall thickness of at least 50%, in order to achieve a radially oriented crystal structure; and annealing the final sized tubing at a temperature and time sufficient to effect complete recrystallization and reform grains in a weld area along the seam into smaller, homogeneous grains.« less

Creep-Rupture Behavior of Ni-Based Alloy Tube Bends for A-USC Boilers

NASA Astrophysics Data System (ADS)

Shingledecker, John

Advanced ultrasupercritical (A-USC) boiler designs will require the use of nickel-based alloys for superheaters and reheaters and thus tube bending will be required. The American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code Section II PG-19 limits the amount of cold-strain for boiler tube bends for austenitic materials. In this summary and analysis of research conducted to date, a number of candidate nickel-based A-USC alloys were evaluated. These alloys include alloy 230, alloy 617, and Inconel 740/740H. Uniaxial creep and novel structural tests and corresponding post-test analysis, which included physical measurements, simplified analytical analysis, and detailed microscopy, showed that different damage mechanisms may operate based on test conditions, alloy, and cold-strain levels. Overall, creep strength and ductility were reduced in all the alloys, but the degree of degradation varied substantially. The results support the current cold-strain limits now incorporated in ASME for these alloys for long-term A-USC boiler service.

Characterization of Tubing from Advanced ODS alloy (FCRD-NFA1)

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

Maloy, Stuart Andrew; Aydogan, Eda; Anderoglu, Osman

2016-09-20

Fabrication methods are being developed and tested for producing fuel clad tubing of the advanced ODS 14YWT and FCRD-NFA1 ferritic alloys. Three fabrication methods were based on plastically deforming a machined thick-wall tube sample of the ODS alloys by pilgering, hydrostatic extrusion or drawing to decrease the outer diameter and wall thickness and increase the length of the final tube. The fourth fabrication method consisted of the additive manufacturing approach involving solid-state spray deposition (SSSD) of ball milled and annealed powder of 14YWT for producing thin-wall tubes. Of the four fabrication methods, two methods were successful at producing tubing formore » further characterization: production of tubing by high-velocity oxy-fuel spray forming and production of tubing using high-temperature hydrostatic extrusion. The characterization described shows through neutron diffraction the texture produced during extrusion while maintaining the beneficial oxide dispersion. In this research, the parameters for innovative thermal spray deposition and hot extrusion processing methods have been developed to produce the final nanostructured ferritic alloy (NFA) tubes having approximately 0.5 mm wall thickness. Effect of different processing routes on texture and grain boundary characteristics has been investigated. It was found that hydrostatic extrusion results in combination of plane strain and shear deformations which generate rolling textures of α- and γ-fibers on {001}<110> and {111}<110> together with a shear texture of ζ-fiber on {011}<211> and {011}<011>. On the other hand, multi-step plane strain deformation in cross directions leads to a strong rolling textures of θ- and ε-fiber on {001}<110> together with weak γ-fiber on {111}<112>. Even though the amount of the equivalent strain is similar, shear deformation leads to much lower texture indexes compared to the plane strain deformations. Moreover, while 50% of hot rolling brings about a large

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.

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

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

Performance assessment of femoral knee components made from cobalt-chromium alloy and oxidized zirconium.

PubMed

Brandt, J-M; Guenther, L; O'Brien, S; Vecherya, A; Turgeon, T R; Bohm, E R

2013-12-01

The surface characteristics of the femoral component affect polyethylene wear in modular total knee replacements. In the present retrieval study, the surface characteristics of cobalt-chromium (CoCr) alloy and oxidized zirconium (OxZr) femoral components were assessed and compared. Twenty-six retrieved CoCr alloy femoral components were matched with twenty-six retrieved OxZr femoral components for implantation period, body-mass index, patient gender, implant type, and polyethylene insert thickness. The surface damage on the retrieved femoral components was evaluated using a semi-quantitative assessment method, scanning electron microscopy, and contact profilometry. The retrieved CoCr alloy femoral components showed less posterior surface gouging than OxZr femoral components; however, at a higher magnification, the grooving damage features on the retrieved CoCr alloy femoral components confirmed an abrasive wear mechanism. The surface roughness values Rp, Rpm, and Rpk for the retrieved CoCr alloy femoral components were found to be significantly higher than those of the retrieved OxZr femoral components (p≤0.031). The surface roughness values were higher on the medial condyles than on the lateral condyles of the retrieved CoCr alloy femoral components; such a difference was not observed on the retrieved OxZr femoral components. The surface roughness of CoCr alloy femoral components increased while the surface roughness of the OxZr femoral components remained unchanged after in vivo service. Therefore, the OxZr femoral components' resistance to abrasive wear may enable lower polyethylene wear and ensure long-term durability in vivo. Level IV. Crown Copyright © 2013. Published by Elsevier B.V. All rights reserved.

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.

Simulation studies on Tube End Expansion of AA2014 Alloy Tubes

NASA Astrophysics Data System (ADS)

Venugopal, L.; Prasad, N. E. C.; Geeta Krishna, P.; Praveen, L.

2018-03-01

End forming is defined as forming the end of tubular forms either by inverting the tube or by expanding it. It finds application in many fields such as in automotive and aerospace sectors as power transmission elements, fuel lines, exhaust pipes etc. The main aim of the present work is to expand the AA2014 alloy tubes with different die sets without any fracture. Deform 2D software was used for performing simulations on expanding the tubes with different die set (punch) values having differed forming angles (α = 15°, 30° and 45°) and expansion ratios (rp/r0 = 1.39, 1.53 and 1.67). Experiments were also conducted and the results correlate with the simulation results. The results shows that for the punch having less cone angle (α) values the linear displacement is more rather than higher cone angles. But in the case of higher cone angles the radial displacement is more than the linear displacement.

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

Evaluation of candidate Stirling engine heater tube alloys after 3500 hours exposure to high pressure doped hydrogen or helium

NASA Technical Reports Server (NTRS)

Misencik, J. A.; Titran, R. H.

1984-01-01

The heater head tubes of current prototype automotive Stirling engines are fabricated from alloy N-155, an alloy which contains 20 percent cobalt. Because the United States imports over 90 percent of the cobalt used in this country and resource supplies could not meet the demand imposed by automotive applications of cobalt in the heater head (tubes plus cylinders and regenerator housings), it is imperative that substitute alloys free of cobalt be identified. The research described herein focused on the heater head tubes. Sixteen alloys (15 potential substitutes plus the 20 percent Co N-155 alloy) were evaluated in the form of thin wall tubing in the NASA Lewis Research Center Stirling simulator materials diesel fuel fired test rigs. Tubes filled with either hydrogen doped with 1 percent CO2 or with helium at a gas pressure of 15 MPa and a temperature of 820 C were cyclic endurance tested for times up to 3500 hr. Results showed that two iron-nickel base superalloys, CG-27 and Pyromet 901 survived the 3500 hr endurance test. The remaining alloys failed by creep-rupture at times less than 3000 hr, however, several other alloys had superior lives to N-155. Results further showed that doping the hydrogen working fluid with 1 vol % CO2 is an effective means of reducing hydrogen permeability through all the alloy tubes investigated.

Effect of oxide films on hydrogen permeability of candidate Stirling heater head tube alloys

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

Schuon, S R; Misencik, J A

1981-01-01

High pressure hydrogen has been selected as the working fluid for the developmental automotive Stirling engine. Containment of the working fluid during operation of the engine at high temperatures and at high hydrogen gas pressures is essential for the acceptance of the Stirling engine as an alternative to the internal combustion engine. Most commercial alloys are extremely permeable to pure hydrogen at high temperatures. A program was undertaken at NASA Lewis Research Center (LeRC) to reduce hydrogen permeability in the Stirling engine heater head tubes by doping the hydrogen working fluid with CO or CO/sub 2/. Small additions of thesemore » gases were shown to form an oxide on the inside tube wall and thus reduce hydrogen permeability. A study of the effects of dopant concentration, alloy composition, and effects of surface oxides on hydrogen permeability in candidate heater head tube alloys is summarized. Results showed that hydrogen permeability was similar for iron-base alloys (N-155, A286, IN800, 19-9DL, and Nitronic 40), cobalt-base alloys (HS-188) and nickel-base alloys (IN718). In general, the permeability of the alloys decreased with increasing concentration of CO or CO/sub 2/ dopant, with increasing oxide thickness, and decreasing oxide porosity. At high levels of dopants, highly permeable liquid oxides formed on those alloys with greater than 50% Fe content. Furthermore, highly reactive minor alloying elements (Ti, Al, Nb, and La) had a strong influence on reducing hydrogen permeability.« 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.

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.

Containerless processing of Nb-Ge alloys in a long drop tube

NASA Technical Reports Server (NTRS)

Bayuzick, R. J.

1982-01-01

The thirty-two meter drop tube at the Marshall Space Flight Center was used to study the effect of zero gravity containerless processing on the structure and properties of materials. The concept involves the suppression of heterogeneous nucleation of solid in liquid and, therefore, solidification accompanied by large degrees of undercooling. Under these conditions metastable phases can be formed or, at the very least, unique nonequilibrium microstructures (containing equilibrium phases) with unique properties can be produced. The drop tube solidification was applied to niobium base alloys with emphasis on the Nb-Ge binary system in an effort to produce metastable phases with high superconducting transition temperatures in bulk specimens. In the past, only lower Ge alloys (Nb-13 a/o, Nb-18 a/o, and Nb-22 a/o) could be undercooled. Higher Ge alloys (e.g., Nb-25 a/o Ge and Nb-27 a/o Ge) can now be undercooled on a routine basis.

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

Cross-Roll Flow Forming of ODS Alloy Heat Exchanger Tubes For Hoop Creep Enhancement

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

Bimal Kad

2007-09-30

Mechanically alloyed oxide dispersion strengthened (ODS) Fe-Cr-Al alloy thin walled tubes and sheets, produced via powder processing and consolidation methodologies are promising materials for eventual use at temperatures up to 1200 C in the power generation industry, far above the temperature capabilities of conventional alloys. Target end-uses range from gas turbine combustor liners to high aspect ratio (L/D) heat exchanger tubes. Grain boundary creep processes at service temperatures, particularly those acting in the hoop direction, are the dominant failure mechanisms for such components. The processed microstructure of ODS alloys consists of high aspect ratio grains aligned parallel to the tubemore » axis, a result of dominant axial metal flow which aligns the dispersoid particles and other impurities in the longitudinal direction. The dispersion distribution is unaltered on a micro scale by recrystallization thermal treatments, but the high aspect ratio grain shape typically obtained limits transverse grain spacing and consequently the hoop creep response. Improving hoop creep in ODS-alloy components will require understanding and manipulating the factors that control the recrystallization behavior, and represents a critical materials design and development challenge that must be overcome in order to fully exploit the potential of ODS alloys. The objectives of this program were to (1) increase creep-strength at temperature in ODS-alloy tube and liner components by 100% via, (2) preferential cross-roll flow forming and grain/particle fibering in the critical hoop direction. The research program outlined was iterative and intended to systematically (i) examine and identify post-extrusion forming methodologies to create hoop strengthened tubes, to be (ii) evaluated at 'in-service' loads at service temperatures and environments. Our report outlines the significant hoop creep enhancements possible via secondary cross-rolling and/or flow-forming operations. Each

Simulating Porous Magnetite Layer Deposited on Alloy 690TT Steam Generator Tubes

PubMed Central

Jeon, Soon-Hyeok; Son, Yeong-Ho; Choi, Won-Ik; Song, Geun Dong; Hur, Do Haeng

2018-01-01

In nuclear power plants, the main corrosion product that is deposited on the outside of steam generator tubes is porous magnetite. The objective of this study was to simulate porous magnetite that is deposited on thermally treated (TT) Alloy 690 steam generator tubes. A magnetite layer was electrodeposited on an Alloy 690TT substrate in an Fe(III)-triethanolamine solution. After electrodeposition, the dense magnetite layer was immersed to simulate porous magnetite deposits in alkaline solution for 50 days at room temperature. The dense morphology of the magnetite layer was changed to a porous structure by reductive dissolution reaction. The simulated porous magnetite layer was compared with flakes of steam generator tubes, which were collected from the secondary water system of a real nuclear power plant during sludge lancing. Possible nuclear research applications using simulated porous magnetite specimens are also proposed. PMID:29301316

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

Production Process of Biocompatible Magnesium Alloy Tubes Using Extrusion and Dieless Drawing Processes

NASA Astrophysics Data System (ADS)

Kustra, Piotr; Milenin, Andrij; Płonka, Bartłomiej; Furushima, Tsuyoshi

2016-06-01

Development of technological production process of biocompatible magnesium tubes for medical applications is the subject of the present paper. The technology consists of two stages—extrusion and dieless drawing process, respectively. Mg alloys for medical applications such as MgCa0.8 are characterized by low technological plasticity during deformation that is why optimization of production parameters is necessary to obtain good quality product. Thus, authors developed yield stress and ductility model for the investigated Mg alloy and then used the numerical simulations to evaluate proper manufacturing conditions. Grid Extrusion3d software developed by authors was used to determine optimum process parameters for extrusion—billet temperature 400 °C and extrusion velocity 1 mm/s. Based on those parameters the tube with external diameter 5 mm without defects was manufactured. Then, commercial Abaqus software was used for modeling dieless drawing. It was shown that the reduction in the area of 60% can be realized for MgCa0.8 magnesium alloy. Tubes with the final diameter of 3 mm were selected as a case study, to present capabilities of proposed processes.

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.

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.

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.

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.

Alloy selection considerations and service experience of the first ''in-service'' 439 stainless steel moisture-separator-reheater tube bundles at Kewaunee nuclear power plant

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

Kratz, J.L.; Minard, P.G.; Weinberg, D.E.

1982-01-01

The need for an alternate moisture-reheater tubing alloy is explored and the reasoning behind the selection of 439 stainless steel is presented. Significant advantages that are evident by using special tubing chemistry and special tubing heat treatment are discussed in relation to fin-die wear, alloy stabilization, maintaining a fully ferritic structure, and reducing the susceptibility to stress-corrosion cracking. Comparisons made between the fatigue response of 439SS tube-to-tube sheet welded specimens ''in air'' at 525/sup 0/F (274/sup 0/C) show a distinct advantage of the use of the 439SS tubing alloy over previously used tubing alloys. An ''in-service'' record of over twomore » years at Kewaunee shows excellent tubing operating experience.« 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.

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.

Fabrication of seamless calandria tubes by cold pilgering route using 3-pass and 2-pass schedules

NASA Astrophysics Data System (ADS)

Saibaba, N.

2008-12-01

Calandria tube is a large diameter, extremely thin walled zirconium alloy tube which has diameter to wall thickness ratio as high as 90-95. Such tubes are conventionally produced by the 'welded route', which involves extrusion of slabs followed by a series of hot and cold rolling passes, intermediate anneals, press forming of sheets into circular shape and closing the gap by TIG welding. Though pilgering is a well established process for the fabrication of seamless tubes, production of extremely thin walled tubes offers several challenges during pilgering. Nuclear fuel complex (NFC), Hyderabad, has successfully developed a process for the production of Zircaloy-4 calandria tubes by adopting the 'seamless route' which involves hot extrusion of mother blanks followed by three-pass pilgering or two-pass pilgering schedules. This paper deals with standardization of the seamless route processes for fabrication of calandria tubes, comparison between the tubes produced by 2-pass and 3-pass pilgering schedules, role of ultrasonic test charts for control of process parameters, development of new testing methods for burst testing and other properties.

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.

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.

Impact Fretting Wear Behavior of Alloy 690 Tubes in Dry and Deionized Water Conditions

NASA Astrophysics Data System (ADS)

Cai, Zhen-Bing; Peng, Jin-Fang; Qian, Hao; Tang, Li-Chen; Zhu, Min-Hao

2017-07-01

The impact fretting wear has largely occurred at nuclear power device induced by the flow-induced vibration, and it will take potential hazards to the service of the equipment. However, the present study focuses on the tangential fretting wear of alloy 690 tubes. Research on impact fretting wear of alloy 690 tubes is limited and the related research is imminent. Therefore, impact fretting wear behavior of alloy 690 tubes against 304 stainless steels is investigated. Deionized water is used to simulate the flow environment of the equipment, and the dry environment is used for comparison. Varied analytical techniques are employed to characterize the wear and tribochemical behavior during impact fretting wear. Characterization results indicate that cracks occur at high impact load in both water and dry equipment; however, the water as a medium can significantly delay the cracking time. The crack propagation behavior shows a jagged shape in the water, but crack extended disorderly in dry equipment because the water changed the stress distribution and retarded the friction heat during the wear process. The SEM and XPS analysis shows that the main failure mechanisms of the tube under impact fretting are fatigue wear and friction oxidation. The effect of medium(water) on fretting wear is revealed, which plays a potential and promising role in the service of nuclear power device and other flow equipments.

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.

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.

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.

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.

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.

SCC analysis of Alloy 600 tubes from a retired steam generator

NASA Astrophysics Data System (ADS)

Hwang, Seong Sik; Kim, Hong Pyo

2013-09-01

Steam generators (SG) equipped with Alloy 600 tubes of a Korean nuclear power plants were replaced with a new one having Alloy 690 tubes in 1998 after 20 years of operation. To set up a guide line for an examination of the other SG tubes, a metallographic examination of the defected tubes was carried out. A destructive analysis on 71 tubes was addressed, and a relation among the stress corrosion crack (SCC) defect location, defect depth, and location of the sludge pile was obtained. Tubes extracted from the retired SG were transferred to a hot laboratory. Detailed nondestructive analysis examinations were taken again at the laboratory, and the tubes were then destructively examined. The types and sizes of the cracks were characterized. The location and depth of the SCC were evaluated in terms of the location and height of the sludge. Most axial cracks were in the sludge pile, whereas the circumferential ones were around the top of the tube sheet (TTS) or below the TTS. Average defect depth of the axial cracks was deeper than that of the circumferential ones. Axial cracks at tube support plate (TSP) seem to be related with corrosion/sludge in crevice like at the TTS region. Circumferential cracks at TSP seem to be caused by tube denting at the upper part of the TSP. Tubes not having clear ECT signals for quantifying an ECT data-base. Tubes having no ECT signal. Tubes with a large ECT signal. Tubes with various types and sizes of flaws (primary water stress corrosion cracking (PWSCC), outside diameter stress corrosion cracking (ODSCC), Pit). Tubes with distinct PWSCC or ODSCC. Tubes were extracted from the RSG based on the field ECT with the criteria, and transferred to a hot laboratory at the Korea Atomic Energy Research Institute (KAERI) for destructive examination. A comprehensive ECT inspection was performed again at the hot laboratory to confirm the location of the cracks obtained from a field inspection. These exact locations of the defects were marked on the

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.

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)

Plate-shaped transformation products in zirconium-base alloys

NASA Astrophysics Data System (ADS)

Banerjee, S.; Dey, G. K.; Srivastava, D.; Ranganathan, S.

1997-11-01

Plate-shaped products resulting from martensitic, diffusional, and mixed mode transformations in zirconium-base alloys are compared in the present study. These alloys are particularly suitable for the comparison in view of the fact that the lattice correspondence between the parent β (bcc) and the product α (hcp) or γ-hydride (fct) phases are remarkably similar for different types of transformations. Crystallographic features such as orientation relations, habit planes, and interface structures associated with these transformations have been compared, with a view toward examining whether the transformation mechanisms have characteristic imprints on these experimental observables. Martensites exhibiting dislocated lath, internally twinned plate, and self-accommodating three-plate cluster morphologies have been encountered in Zr-2.5Nb alloy. Habit planes corresponding to all these morphologies have been found to be consistent with the predictions based on the invariant plane strain (IPS) criterion. Different morphologies have been found to reflect the manner in which the neighboring martensite variants are assembled. Lattice-invariant shears (LISs) for all these cases have been identified to be either {10 bar 11} α < bar 1123> α slip or twinning on {10 bar 11} α planes. Widmanstätten α precipitates, forming in a step-quenching treatment, have been shown to have a lath morphology, the α/β interface being decorated with a periodic array of < c + a> dislocations at a spacing of 8 to 10 nm. The line vectors of these dislocations are nearly parallel to the invariant lines. The α precipitates, forming in the retained β phase on aging, exhibit an internally twinned structure with a zigzag habit plane. Average habit planes for the morphologies have been found to lie near the {103} β — {113} β poles, which are close to the specific variant of the {112} β plane, which transforms into a prismatic plane of the type {1 bar 100} α . The crystallography of the

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.

One Dimensional Cold Rolling Effects on Stress Corrosion Crack Growth in Alloy 690 Tubing and Plate Materials

NASA Astrophysics Data System (ADS)

Toloczko, Mychailo B.; Olszta, Matthew J.; Bruemmer, Stephen M.

Stress corrosion crack-growth experiments have been performed on cold-rolled alloy 690 materials in simulated PWR primary water at 360°C. Extruded alloy 690 CRDM tubing in two conditions, thermally treated (TT) and solution annealed (SA), was cold rolled (CR) in one direction to several reductions reaching a maximum of 31% and tested in the S-L orientation. High stress corrosion cracking (SCC) propagation rates ( 8x10-8 mm/s) were observed for the 31%CR alloy 690TT material, while the 31%CR alloy 690SA exhibited 10X lower rates. The difference in intergranular SCC susceptibility appears to be related to grain boundary carbide distribution before cold rolling. SCC growth rates were found to depend on test temperature and hydrogen concentration. Tests were also performed on two alloy 690 plate heats, one CR to a reduction of 26% and the other to 20%. SCC growth rates at 360°C were similar to that measured for the 31%CR alloy 690TT CRDM tubing. Comparisons will be made to other results on CR alloy 690 materials.

Optimized manufacture of nuclear fuel cladding tubes by FEA of hot extrusion and cold pilgering processes

NASA Astrophysics Data System (ADS)

Gaillac, Alexis; Ly, Céline

2018-05-01

Within the forming route of Zirconium alloy cladding tubes, hot extrusion is used to deform the forged billets into tube hollows, which are then cold rolled to produce the final tubes with the suitable properties for in-reactor use. The hot extrusion goals are to give the appropriate geometry for cold pilgering, without creating surface defects and microstructural heterogeneities which are detrimental for subsequent rolling. In order to ensure a good quality of the tube hollows, hot extrusion parameters have to be carefully chosen. For this purpose, finite element models are used in addition to experimental tests. These models can take into account the thermo-mechanical coupling conditions obtained in the tube and the tools during extrusion, and provide a good prediction of the extrusion load and the thermo-mechanical history of the extruded product. This last result can be used to calculate the fragmentation of the microstructure in the die and the meta-dynamic recrystallization after extrusion. To further optimize the manufacturing route, a numerical model of the cold pilgering process is also applied, taking into account the complex geometry of the tools and the pseudo-steady state rolling sequence of this incremental forming process. The strain and stress history of the tube during rolling can then be used to assess the damage risk thanks to the use of ductile damage models. Once validated vs. experimental data, both numerical models were used to optimize the manufacturing route and the quality of zirconium cladding tubes. This goal was achieved by selecting hot extrusion parameters giving better recrystallized microstructure that improves the subsequent formability. Cold pilgering parameters were also optimized in order to reduce the potential ductile damage in the cold rolled tubes.

Oxidation and corrosion resistance of candidate Stirling engine heater-head-tube alloys

NASA Technical Reports Server (NTRS)

Stephens, J. R.; Barrett, C. A.

1984-01-01

Sixteen candidate iron base Stirling engine heater head tube alloys are evaluated in a diesel fuel fired simulator materials test rig to determine their oxidation and corrosion resistance. Sheet specimens are tested at 820 C for 3500 hr in 5 hr heating cycles. Specific weight change data and an attack parameter are used to categorize the alloys into four groups; 10 alloys show excellent for good oxidation and corrosion resistance and six alloys exhibit poor or catastrophic resistance. Metallographic, X-ray, and electron microprobe analyses aid in further characterizing the oxidation and corrosion behavior of the alloys. Alloy compositions, expecially the reactive elements aluminum, titanium, and chromium, play a major role in the excellent oxidation and corrosion behavior of the alloys. The best oxidation resistance is associated with the formation of an iron nickel aluminum outer oxide scale, an intermediate oxide scale rich in chromium and titanium, and an aluminum outer oxide scale adjacent to the metallic substrate, which exhibits a zone of internal oxidation of aluminum and to some extent titanium.

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.

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.

High strength Sn-Mo-Nb-Zr alloy tubes and method of making same

DOEpatents

Cheadle, Brian A.

1977-01-01

Tubes for use in nuclear reactors fabricated from a quaternary alloy comprising 2.5-4.0 wt% Sn, 0.5-1.5 wt% Mo, 0.5-1.5 wt% Nb, balance essentially Zr. The tubes are fabricated by a process of hot extrusion, heat treatment, cold working to size and age hardening, so as to produce a microstructure comprising elongated .alpha. grains with an acicular transformed .beta. grain boundary phase.

The influence of manufacturing processes on the microstructure, grain boundary characteristics and SCC behavior of Alloy 690 steam generator tubing

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

Sarver, J.M.; Doherty, P.E.; Doyle, D.M.

1995-12-31

Thermally treated Alloy 690 is the tubing material of choice for replacement steam generators in the United States. Throughout the world, it is manufactured using different melting and thermomechanical processing methods. The influence of different processing steps on the intergranular stress corrosion cracking (IGSCC) behavior of Alloy 690 has not been thoroughly evaluated. Evaluations were performed on Alloy 690 steam generator tubing produced using several different melting practices and thermomechanical processing procedures. The evaluations included extensive microstructural examinations as well as constant extension rate (CERT) tests. The CERT test results indicated that the thermally treated Alloy 690 tubing which wasmore » subjected to higher annealing temperatures displayed the highest degree of resistance to stress corrosion cracking (SCC). Examination of the microstructures indicated that the microstructural changes which are produced by increased annealing temperatures are subtle. In an attempt to further elucidate and quantify the effect of manufacturing processes on corrosion behavior, grain boundary character distribution (GBCD) measurements were performed on the same materials which were CERT tested. Analysis of GBCDs of the samples used in this study indicate that Alloy 690 exhibits a significantly larger fraction of special boundaries as compared to Alloy 600 and Alloy 800, regardless of the processing history of the tubing. Preliminary results indicate that a correlation may exist between processing method, GBCD`s and degree of IGSCC exhibited by the thermally treated samples examined in this study.« less

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

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

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.

Fatigue of Ti-3A1-2.5V Alloy Tube and Rod

DTIC Science & Technology

2007-02-28

during cyclic straining of a 18% nickel maraging steel and attributed it to the presence of a crack. Skelton (reference 6) also attributed a hysteresis...growth, and fracture lives was also defined. The LCF behavior of the alloy tube and rod was investigated, examining the shape change of load...is weldable as the commercially pure grades and has excellent resistance to torsion and corrosion . Therefore, it is used principally as tubing in

Inconel 690 is alloy of choice for steam-generator tubing

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

Strauss, S.D.

1996-02-01

The product of two decades of research and plant application, Inconel 690 promises superior long-term resistance to tube cracking in comparison to alloy 600. Ongoing steam-generator management techniques applied at nuclear pressurized-water-reactor (PWR) plants focus on tube monitoring, inspection, and repair, and on water-chemistry control. Of greatest concern to owner/operators of steam generators (SGs) with recirculating (U-bend) rather than straight through tubes is corrosion of several forms, including pitting, thinning, and cracking. As problems persist and operating and maintenance (O and M) costs become prohibitive, managers must consider the remaining option: complete or partial SG replacement. Although replacement costs canmore » range upward of $100-million, this step restores full-power operation, simplifies inspection, shortens subsequent outages, increases unit availability, and reduces radiation exposure of maintenance personnel. Taken together, these can lead to economies over the long term.« less

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

Surface treatment to form a dispersed Y2O3 layer on Zircaloy-4 tubes

NASA Astrophysics Data System (ADS)

Jung, Yang-Il; Kim, Hyun-Gil; Guim, Hwan-Uk; Lim, Yoon-Soo; Park, Jung-Hwan; Park, Dong-Jun; Yang, Jae-Ho

2018-01-01

Zircaloy-4 is a traditional zirconium-based alloy developed for application in nuclear fuel cladding tubes. The surfaces of Zircaloy-4 tubes were treated using a laser beam to increase their mechanical strength. Laser beam scanning of a tube coated with yttrium oxide (Y2O3) resulted in the formation of a dispersed oxide layer in the tube's surface region. Y2O3 particles penetrated the Zircaloy-4 during the laser treatment and were distributed uniformly in the surface region. The thickness of the dispersed oxide layer varied from 50 to 140 μm depending on the laser beam trajectory. The laser treatment also modified the texture of the tube. The preferred basal orientation along the normal to the tube surface disappeared, and a random structure appeared after laser processing. The most obvious result was an increase in the mechanical strength. The tensile strength of Zircaloy-4 increased by 10-20% with the formation of the dispersed oxide layer. The compressive yield stress also increased, by more than 15%. Brittle fracture was observed in the surface-treated samples during tensile and compressive deformation at room temperature; however, the fracture behavior was changed in ductile at elevated temperatures.

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.

Effect of oxide films on hydrogen permeability of candidate Stirling engine heater head tube alloys

NASA Technical Reports Server (NTRS)

Schuon, S. R.; Misencik, J. A.

1981-01-01

The effect of oxide films developed in situ from CO/CO2 doped hydrogen on high pressure hydrogen permeability at 820 C was studied on N-155, A-286, IN 800, 19-9DL, Nitronic 40, HS-188, and IN 718 tubing in a Stirling materials simulator. The hydrogen permeability decreased with increasing dopant levels of CO or CO2 and corresponding decreases in oxide porosity. Minor reactive alloying elements strongly influenced permeability. At high levels of CO or CO2, a liquid oxide formed on alloys with greater than 50 percent Fe. This caused increased permeability. The oxides formed on the inside tube walls were analyzed and their effective permeabilities were calculated.

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.

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.

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

Fabrication of Aluminum Tubes Filled with Aluminum Alloy Foam by Friction Welding.

PubMed

Hangai, Yoshihiko; Nakano, Yukiko; Koyama, Shinji; Kuwazuru, Osamu; Kitahara, Soichiro; Yoshikawa, Nobuhiro

2015-10-23

Aluminum foam is usually used as the core of composite materials by combining it with dense materials, such as in Al foam core sandwich panels and Al-foam-filled tubes, owing to its low tensile and bending strengths. In this study, all-Al foam-filled tubes consisting of ADC12 Al-Si-Cu die-cast aluminum alloy foam and a dense A1050 commercially pure Al tube with metal bonding were fabricated by friction welding. First, it was found that the ADC12 precursor was firmly bonded throughout the inner wall of the A1050 tube without a gap between the precursor and the tube by friction welding. No deformation of the tube or foaming of the precursor was observed during the friction welding. Next, it was shown that by heat treatment of an ADC12-precursor-bonded A1050 tube, gases generated by the decomposition of the blowing agent expand the softened ADC12 to produce the ADC12 foam interior of the dense A1050 tube. A holding time during the foaming process of approximately t H = 8.5 min with a holding temperature of 948 K was found to be suitable for obtaining a sound ADC12-foam-filled A1050 tube with sufficient foaming, almost uniform pore structures over the entire specimen, and no deformation or reduction in the thickness of the tube.

Fabrication of Aluminum Tubes Filled with Aluminum Alloy Foam by Friction Welding

PubMed Central

Hangai, Yoshihiko; Nakano, Yukiko; Koyama, Shinji; Kuwazuru, Osamu; Kitahara, Soichiro; Yoshikawa, Nobuhiro

2015-01-01

Aluminum foam is usually used as the core of composite materials by combining it with dense materials, such as in Al foam core sandwich panels and Al-foam-filled tubes, owing to its low tensile and bending strengths. In this study, all-Al foam-filled tubes consisting of ADC12 Al-Si-Cu die-cast aluminum alloy foam and a dense A1050 commercially pure Al tube with metal bonding were fabricated by friction welding. First, it was found that the ADC12 precursor was firmly bonded throughout the inner wall of the A1050 tube without a gap between the precursor and the tube by friction welding. No deformation of the tube or foaming of the precursor was observed during the friction welding. Next, it was shown that by heat treatment of an ADC12-precursor-bonded A1050 tube, gases generated by the decomposition of the blowing agent expand the softened ADC12 to produce the ADC12 foam interior of the dense A1050 tube. A holding time during the foaming process of approximately tH = 8.5 min with a holding temperature of 948 K was found to be suitable for obtaining a sound ADC12-foam-filled A1050 tube with sufficient foaming, almost uniform pore structures over the entire specimen, and no deformation and minimum reduction in the thickness of the tube. PMID:28793629

Brazing open cell reticulated copper foam to stainless steel tubing with vacuum furnace brazed gold/indium alloy plating

DOEpatents

Howard, Stanley R [Windsor, SC; Korinko, Paul S [Aiken, SC

2008-05-27

A method of fabricating a heat exchanger includes brush electroplating plated layers for a brazing alloy onto a stainless steel tube in thin layers, over a nickel strike having a 1.3 .mu.m thickness. The resultant Au-18 In composition may be applied as a first layer of indium, 1.47 .mu.m thick, and a second layer of gold, 2.54 .mu.m thick. The order of plating helps control brazing erosion. Excessive amounts of brazing material are avoided by controlling the electroplating process. The reticulated copper foam rings are interference fit to the stainless steel tube, and in contact with the plated layers. The copper foam rings, the plated layers for brazing alloy, and the stainless steel tube are heated and cooled in a vacuum furnace at controlled rates, forming a bond of the copper foam rings to the stainless steel tube that improves heat transfer between the tube and the copper foam.

Compression Behavior and Energy Absorption of Aluminum Alloy AA6061 Tubes with Multiple Holes

NASA Astrophysics Data System (ADS)

Simhachalam, Bade; Lakshmana Rao, C.; Srinivas, Krishna

2014-05-01

In this article, compression behavior and energy absorption of aluminum alloy AA6061 tubes are investigated both experimentally and numerically. Static and dynamic simulations are done using LS-Dyna Software for AA6061 tubes. True stress-plastic strain curves from the tensile test are used in the static and dynamic simulations of AA6061 tubes. The energy absorption values between experimental compression results and numeral simulation are found to be in good agreement. Dynamic simulations are done with drop velocity of up to 10 m/s to understand the inertia effects on energy absorption. The deformed modes from the numerical simulation are compared between tubes with and without holes in static and dynamic conditions.

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

Study on Hydroforming of Magnesium Alloy Tube under Temperature Condition

NASA Astrophysics Data System (ADS)

Wang, Xinsong; Wang, Shouren; Zhang, Yongliang; Wang, Gaoqi; Guo, Peiquan; Qiao, Yang

2018-01-01

First of all, under 100 °C, 150 °C, 200 °C, 250 °C, 300 °C and 350 °C, respectively do the test of magnesium alloy AZ31B temperature tensile and the fracture of SEM electron microscopic scanning, studying the plastic forming ability under six different temperature. Secondly, observe and study the real stress-strain curves and fracture topography. Through observation and research can concluded that with the increase of temperature, the yield strength and tensile strength of AZ31B was increased, and the elongation rate and the plastic deformation capacity are increased obviously. Taking into account the actual production, energy consumption, and mold temperature resistance, 250 °Cwas the best molding temperature. Finally, under the temperature condition of 250 °C, the finite element simulation and simulation of magnesium alloy profiled tube were carried out by Dynaform, and the special wall and forming limit diagram of magnesium alloy were obtained. According to the forming wall thickness and forming limit diagram, the molding experiment can be optimized continuously.

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.

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.

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.

Proving the viability of manufacturing of multi-layer steel/vanadium alloy/steel composite tubes by numerical simulations and experiment

NASA Astrophysics Data System (ADS)

Nechaykina, T.; Nikulin, S.; Rozhnov, A.; Molotnikov, A.; Zavodchikov, S.; Estrin, Y.

2018-05-01

Vanadium alloys are promising structural materials for fuel cladding tubes for fast-neutron reactors. However, high solubility of oxygen and nitrogen in vanadium alloys at operating temperatures of 700 °C limits their application. In this work, we present a novel composite structure consisting of vanadium alloy V-4Ti-4Cr (provides high long-term strength of the material) and stainless steel Fe-0.2C-13Cr (as a corrosion resistant protective layer). It is produced by co-extrusion of these materials forming a three-layered tube. Finite element simulations were utilised to explore the influence of the various co-extrusion parameters on manufacturability of multi-layered tubes. Experimental verification of the numerical modelling was performed using co-extrusion with the process parameters suggested by the numerical simulations. Scanning electron microscopy and microhardness measurements revealed a defect-free diffusion layer at the interfaces between both materials indicating a good quality bonding for these co-extrusion conditions.

Study of IGA/SCC behavior of alloy 600 and 690 SG tubing materials in high temperature solutions

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

Tsujikawa, S.; Yashima, S.; Hattori, T.

1996-09-01

Intergranular attack/stress corrosion cracking (IGA/SCC) of Alloy 600 Steam Generator (SG) tubes in the secondary side has been recognized as a matter of great concern for PWRs. Here, IGA/SCC behavior of Alloy 600 and 690 in high temperature solutions was studied using constant extension rate testing (CERT) method under potentiostatic conditions. The IGA/SCC susceptible regions were investigated as a function of pH and electrode potential. The IGA/SCC resistance of SG tubing materials were ranked as, MA600 = TT600 {much_lt} TT690 in acidic solutions, and MA600 < TT600 < TT690 in alkaline solutions. TT690 showed higher corrosion resistance than MA600 andmore » TT600 in both acidic and alkaline conditions. To verify the results of CERT test, long term model boiler tests were also carried out. The model boiler which consists of combinations of several SG tubing materials and tube support plate configurations, operated for more than 15,000 hrs under the simulated operating plant conditions. The results of destructive examination showed good correspondence with the results of a fundamental study, CERT test. The improved performance of alternate SG tubing material was confirmed.« less

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.

Evaluation of dopants in hydrogen to reduce hydrogen permeation in candidate Stirling engine heater head tube alloys at 760 deg and 820 deg

NASA Technical Reports Server (NTRS)

Misencik, J. A.

1982-01-01

Alloy tubes filled with hydrogen doped with various amounts of carbon monoxide, carbon dioxide, ethane, ethylene, methane, ammonia, or water were heated in a diesel fuel-fired Stirling engine simulator materials test rig for 100 hours at 21 MPa and 760 or 820 C to determine the effectiveness of the dopants in reducing hydrogen permeation through the hot tube walls. Ultra high purity (UHP) hydrogen was used for comparison. The tube alloys were N-155, A-286, Incoloy 800, Nitronic 40, 19-9DL, 316 stainless steel, Inconel 718, and HS-188. Carbon dioxide and carbon monoxide in the concentration range 0.2 to 5 vol % were most effective in reducing hydrogen permeation through the hot tube walls for all alloys. Ethane, ethylene, methane, ammonia, and water at the concentrations investigated were not effective in reducing the permeation below that achieved with UHP hydrogen. One series of tests were conducted with UHP hydrogen in carburized tubes. Carburization of the tubes prior to exposure reduced permeation to values similar to those for carbon monoxide; however, carbon dioxide was the most effective dopant.

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

An Investigation on Axial Deformation Behavior of Thin-Wall Unfilled and Filled Tube with Aluminum Alloy (Al-Si7Mg) Foam Reinforced with SiC Particles

NASA Astrophysics Data System (ADS)

Kumaraswamidhas, L. A.; Rajak, Dipen Kumar; Das, S.

2016-08-01

The objective of this research is to produce superior quality aluminum alloy foam with low relative density and higher resistance against compression deformation. This investigation has studied crash energy capacities of unfilled and filled aluminum alloy foams in mild steel tubes. The foam has been prepared by the melt route process with an addition of 5wt.% silicon carbide particles. The fabricated aluminum alloy foams were characterized by field emission scanning electron microscopy, x-ray diffraction, Fourier transform infrared spectroscopy, and Material Pro analyzer. It was observed that the foam-filled tubes could absorb more energy as compared to the unfilled tubes before reaching the complete densification point. Also, the aluminum alloy foams had better energy absorption capacity during the crash or impact loading. This article demonstrates the excellent ability of aluminum alloy foam application in the field where there is a need to absorb crash energy. It is to be noted that the amount of energy absorption will be greater for low-density foam filled in thin-wall rectangular section tubes. We have seen an increasing trend in the application of aluminum foams inside the thin-wall mild steel tubes for maximum energy absorption.

Deformation Mechanisms in Tube Billets from Zr-1%Nb Alloy under Radial Forging

NASA Astrophysics Data System (ADS)

Perlovich, Yuriy; Isaenkova, Margarita; Fesenko, Vladimir; Krymskaya, Olga; Zavodchikov, Alexander

2011-05-01

Features of the deformation process by cold radial forging of tube billets from Zr-1%Nb alloy were reconstructed on the basis of X-ray data concerning their structure and texture. The cold radial forging intensifies grain fragmentation in the bulk of billet and increases significantly the latent hardening of potentially active slip systems, so that operation only of the single slip system becomes possible. As a result, in radially-forged billets unusual deformation and recrystallization textures arise. These textures differ from usual textures of α-Zr by the mutual inversion of crystallographic axes, aligned along the axis of tube.

High rate constitutive modeling of aluminium alloy tube

NASA Astrophysics Data System (ADS)

Salisbury, C. P.; Worswick, M. J.; Mayer, R.

2006-08-01

As the need for fuel efficient automobiles increases, car designers are investigating light-weight materials for automotive bodies that will reduce the overall automobile weight. Aluminium alloy tube is a desirable material to use in automotive bodies due to its light weight. However, aluminium suffers from lower formability than steel and its energy absorption ability in a crash event after a forming operation is largely unknown. As part of a larger study on the relationship between crashworthiness and forming processes, constitutive models for 3mm AA5754 aluminium tube were developed. A nominal strain rate of 100/s is often used to characterize overall automobile crash events, whereas strain rates on the order of 1000/s can occur locally. Therefore, tests were performed at quasi-static rates using an Instron test fixture and at strain rates of 500/s to 1500/s using a tensile split Hopkinson bar. High rate testing was then conducted at rates of 500/s, 1000/s and 1500/s at 21circC, 150circC and 300circC. The generated data was then used to determine the constitutive parameters for the Johnson-Cook and Zerilli-Armstrong material models.

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.

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.

Effect of hydrogenation conditions on the microstructure and mechanical properties of zirconium hydride

NASA Astrophysics Data System (ADS)

Muta, Hiroaki; Nishikane, Ryoji; Ando, Yusuke; Matsunaga, Junji; Sakamoto, Kan; Harjo, Stefanus; Kawasaki, Takuro; Ohishi, Yuji; Kurosaki, Ken; Yamanaka, Shinsuke

2018-03-01

Precipitation of brittle zirconium hydrides deteriorate the fracture toughness of the fuel cladding tubes of light water reactor. Although the hydride embrittlement has been studied extensively, little is known about physical properties of the hydride due to the experimental difficulties. In the present study, to elucidate relationship between mechanical properties and microstructure, two δ-phase zirconium hydrides and one ε-phase zirconium hydride were carefully fabricated considering volume changes at the metal-to-hydride transformation. The δ-hydride that was fabricated from α-zirconium exhibits numerous inner cracks due to the large volume change. Analyses of the neutron diffraction pattern and electron backscatter diffraction (EBSD) data show that the sample displays significant stacking faults in the {111} plane and in the pseudo-layered microstructure. On the other hand, the δ-hydride sample fabricated from β-zirconium at a higher temperature displays equiaxed grains and no cracks. The strong crystal orientation dependence of mechanical properties were confirmed by indentation test and EBSD observation. The δ-hydride hydrogenated from α-zirconium displays a lower Young's modulus than that prepared from β-zirconium. The difference is attributed to stacking faults within the {111} plane, for which the Young's modulus exhibits the highest value in the perpendicular direction. The strong influence of the crystal orientation and dislocation density on the mechanical properties should be considered when evaluating hydride precipitates in nuclear fuel cladding.

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

Torsion Tests of 24S-T Aluminum-alloy Noncircular Bar and Tubing

NASA Technical Reports Server (NTRS)

Moore, R L; Paul, D A

1943-01-01

Tests of 24S-T aluminum alloy have been made to determine the yield and ultimate strengths in torsion of noncircular bar and tubing. An approximate basis for predicting these torsional strength characteristics has been indicated. The results show that the torsional stiffness and maximum shearing stresses within the elastic range may be computed quite closely by means of existing formulas based on mathematical analysis and the membrane analogy.

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

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.

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

Surface Roughening Behavior of 6063 Aluminum Alloy during Bulging by Spun Tubes

PubMed Central

Cai, Yang; Wang, Xiaosong; Yuan, Shijian

2017-01-01

Severe surface roughening during the hydroforming of aluminum alloy parts can produce surface defects that severely restrict their application in the automobile and aerospace industry. To understand the relation between strain, grain size and surface roughness under biaxial stress conditions, hydro-bulging tests of aluminum alloy tubes were carried out, and the tubes with different grain sizes were prepared by a spinning and annealing process. The surface roughness was measured by a laser scanning confocal microscope to evaluate the surface roughening macroscopical behavior, and the corresponding microstructures were observed using electron back-scattered diffraction (EBSD) to reveal the roughening microscopic behavior. The results obtained show that the surface roughness increased with both strain and grain size under biaxial stress. No surface defects were observed on the surface when the grain size was less than 105 μm if the strain was less than 18%, or when the grain size was between 130 and 175 μm if the strain was less than 15.88% and 7.15%, respectively. The surface roughening microscopic behavior was identified as an inhomogeneous grain size distribution, which became more pronounced with increasing grain size and resulted in greater local deformation. Concentrated grain orientation also results in severe inhomogeneous deformation during plastics deformation, and serious surface roughening. PMID:28772658

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.

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).

Surface preparation of Ti-3Al-2.5V alloy tubes for welding using a fiber laser

NASA Astrophysics Data System (ADS)

Kumar, Aniruddha; Gupta, Mool C.

2009-11-01

Ti-3Al-2.5V tubes are widely used in aircraft hydraulic systems. Meticulous surface preparation before welding is necessary to obtain a sound weld involving these alloy tubes. Conventionally this is done by cleaning with environmentally malign toxic chemicals, such as, hydrofluoric acid and nitric acid. This paper describes the laser-cleaning process of the surface of these tubes with a fiber laser as a preparation for pulsed gas tungsten arc welding and results obtained. A simple one-dimensional heat equation has been solved to evaluate the temperature profile of the irradiated surface. It is shown that surface preparation by laser cleaning can be an environmentally friendly alternative process by producing acceptable welds with laser-processed tubes.

The Bulging Behavior of Thick-Walled 6063 Aluminum Alloy Tubes Under Double-Sided Pressures

NASA Astrophysics Data System (ADS)

Cui, Xiao-Lei; Wang, Xiao-Song; Yuan, Shi-Jian

2015-05-01

To make further exploration on the deformation behavior of tube under double-sided pressures, the thick-walled 6063 aluminum alloy tubes with an outer diameter of 65 mm and an average thickness of 7.86 mm have been used to be bulged under the combined action of internal and external pressures. In the experiment, two ends of the thick-walled tubes were fixed using the tooth and groove match. Three levels of external pressure (0 MPa, 40 MPa, and 80 MPa), in conjunction with the internal pressure, were applied on the tube outside and inside simultaneously. The effect of external pressure on the bulging behavior of the thick-walled tubes, such as the limiting expansion ratio, the bulging zone profile, and the thickness distribution, has been investigated. It is shown that the limiting expansion ratio, the bulging zone profile, and the thickness distribution in the homogeneous bulging area are all insensitive to the external pressure. However, the external pressure can make the thick-walled tube achieve a thinner wall at the fracture area. It reveals that the external pressure can only improve the fracture limit of the thick-walled 6063 tubes, but it has very little effect on their homogeneous bulging behavior. It might be because the external pressure can only increase the magnitude of the hydrostatic pressure for the tube but has no effect on the Lode parameter.

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

Demonstration of a shape memory alloy torque tube-based morphing radiator

NASA Astrophysics Data System (ADS)

Chong, Jorge B.; Walgren, Patrick; Hartl, Darren J.

2018-03-01

Long-distance crewed space exploration will require advanced thermal control systems (TCS) with the ability to handle a wide range of thermal loads. The ability of a TCS to adapt to the thermal environment is described by the turndown ratio. Developing radiators with high turndown ratios is critical for improving TCS technology. This paper describes a novel morphing radiator designed to achieve a high turndown ratio by varying its own radiative view factor and effective emissivity through the use of shape memory alloys (SMAs). This radiator features two SMA torque tubes cantilevered to a rigid fixture. The working fluid is transported within the SMA tubes through an annular flow system. In a cold environment, radiator panels fixed to the free ends of the tubes are oriented vertically in a parallel-plate fashion, where the high-emissivity interior faces have restricted views to the environment and heat rejection is minimized. When the system heats up, the tubes actuate by twisting in opposing directions, bringing the panels to a horizontal position with the interior faces exposed to maximize heat rejection. When the system cools down, the tubes twist in reverse, restoring the panels to the vertical orientation where heat rejection is again minimized. This variable heat rejection system has the potential for achieving higher turndown ratios than those of current state-of-the-art systems. A benchtop prototype has been designed and tested to demonstrate actuation and to explore internal heat transfer effects. Prototype design, testing, and results are herein described.

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.

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.

Viability of thin wall tube forming of ATF FeCrAl

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

Maloy, Stuart Andrew; Aydogan, Eda; Anderoglu, Osman

Fabrication of thin walled tubing of FeCrAl alloys is critical to its success as a candidate enhanced accident-tolerant fuel cladding material. Alloys that are being investigated are Generation I and Generation II FeCrAl alloys produced at ORNL and an ODS FeCrAl alloy, MA-956 produced by Special Metals. Gen I and Gen II FeCrAl alloys were provided by ORNL and MA-956 was provided by LANL (initially produced by Special Metals). Three tube development efforts were undertaken. ORNL led the FeCrAl Gen I and Gen II alloy development and tube processing studies through drawing tubes at Rhenium Corporation. LANL received alloys frommore » ORNL and led tube processing studies through drawing tubes at Century Tubing. PNNL led the development of tube processing studies on MA-956 through pilger processing working with Sandvik Corporation. A summary of the recent progress on tube development is provided in the following report and a separate ORNL report: ORNL/TM-2015/478, “Development and Quality Assessments of Commercial Heat Production of ATF FeCrAl Tubes”.« less

Lateral Compression Properties of Magnesium Alloy Tubes Fabricated via Hydrostatic Extrusion Integrated with Circular ECAP

NASA Astrophysics Data System (ADS)

Lv, Jiuming; Hu, Fangyi; Cao, Quoc Dinh; Yuan, Renshu; Wu, Zhilin; Cai, Hongming; Zhao, Lei; Zhang, Xinping

2017-03-01

Hydrostatic extrusion integrated with circular equal channel angular pressing has been previously proposed for fabricating AZ80 magnesium alloy tubes as a method to obtain high-strength tubes for industrial applications. In order to axial tensile strength, circumferential mechanical properties are also important for tubular structures. The tensile properties of AZ80 tubes have been previously studied; however, the circumferential properties have not been examined. In this work, circumferential mechanical properties of these tubes were studied using lateral compression tests. An analytical model is proposed to evaluate the circumferential elongation, which is in good agreement with finite element results. The effects of the extrusion ratio and conical mandrel angle on the circumferential elongation and lateral compression strength are discussed. The strain distribution in the sample during lateral compression testing was found to be inhomogeneous, and cracks initially appeared on the inner surface of the sample vertex. The circumferential elongation and lateral compression strength increased with the extrusion ratio and conical mandrel angle. The anisotropy of the tube's mechanical properties was insignificant when geometric effects were ignored.

Effect of tube processing methods on microstructure, mechanical properties and irradiation response of 14YWT nanostructured ferritic alloys

DOE PAGES

Aydogan, E.; Maloy, S. A.; Anderoglu, O.; ...

2017-06-06

In this research, innovative thermal spray deposition (Process I) and conventional hot extrusion processing (Process II) methods have been used to produce thin walled tubing (~0.5 mm wall thickness) out of 14YWT, a nanostructured ferritic alloy. The effects of processing methods on the microstructure, mechanical properties and irradiation response have been investigated by using scanning electron microscopy (SEM), transmission electron microscopy (TEM) and, micro- and nano-hardness techniques. It has been found that these two processes have a significant effect on the microstructure and mechanical properties of the as-fabricated 14YWT tubes. Even though both processing methods yield the formation of variousmore » size Y-Ti-O particles, the conventional hot extrusion method results in a microstructure with smaller, homogenously distributed nano-oxides (NOs, Y-Ti-O particles < 5 nm) with higher density. Therefore, Process II tubes exhibit twice the hardness of Process I tubes. It has also been found that these two tremendously different initial microstructures strongly affect irradiation response in these tubes under extremely high dose ion irradiations up to 1100 peak dpa at 450 °C. The finer, denser and homogenously distributed NOs in the Process II tube result in a reduction in swelling by two orders of magnitude. On the other hand, inhomogeneity of the initial microstructure in the Process I tube leads to large variations in both swelling and irradiation induced hardening. Moreover, hardening mechanisms before and after irradiation were measured and compared with detailed calculations. In conclusion, this study clearly indicates the crucial effect of initial microstructure on radiation response of 14YWT alloys.« less

Effect of tube processing methods on microstructure, mechanical properties and irradiation response of 14YWT nanostructured ferritic alloys

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

Aydogan, E.; Maloy, S. A.; Anderoglu, O.

In this research, innovative thermal spray deposition (Process I) and conventional hot extrusion processing (Process II) methods have been used to produce thin walled tubing (~0.5 mm wall thickness) out of 14YWT, a nanostructured ferritic alloy. The effects of processing methods on the microstructure, mechanical properties and irradiation response have been investigated by using scanning electron microscopy (SEM), transmission electron microscopy (TEM) and, micro- and nano-hardness techniques. It has been found that these two processes have a significant effect on the microstructure and mechanical properties of the as-fabricated 14YWT tubes. Even though both processing methods yield the formation of variousmore » size Y-Ti-O particles, the conventional hot extrusion method results in a microstructure with smaller, homogenously distributed nano-oxides (NOs, Y-Ti-O particles < 5 nm) with higher density. Therefore, Process II tubes exhibit twice the hardness of Process I tubes. It has also been found that these two tremendously different initial microstructures strongly affect irradiation response in these tubes under extremely high dose ion irradiations up to 1100 peak dpa at 450 °C. The finer, denser and homogenously distributed NOs in the Process II tube result in a reduction in swelling by two orders of magnitude. On the other hand, inhomogeneity of the initial microstructure in the Process I tube leads to large variations in both swelling and irradiation induced hardening. Moreover, hardening mechanisms before and after irradiation were measured and compared with detailed calculations. In conclusion, this study clearly indicates the crucial effect of initial microstructure on radiation response of 14YWT alloys.« less

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

Zirconium and hafnium

USGS Publications Warehouse

Jones, James V.; Piatak, Nadine M.; Bedinger, George M.; Schulz, Klaus J.; DeYoung,, John H.; Seal, Robert R.; Bradley, Dwight C.

2017-12-19

Zirconium and hafnium are corrosion-resistant metals that are widely used in the chemical and nuclear industries. Most zirconium is consumed in the form of the main ore mineral zircon (ZrSiO4, or as zirconium oxide or other zirconium chemicals. Zirconium and hafnium are both refractory lithophile elements that have nearly identical charge, ionic radii, and ionic potentials. As a result, their geochemical behavior is generally similar. Both elements are classified as incompatible because they have physical and crystallochemical properties that exclude them from the crystal lattices of most rock-forming minerals. Zircon and another, less common, ore mineral, baddeleyite (ZrO2), form primarily as accessory minerals in igneous rocks. The presence and abundance of these ore minerals in igneous rocks are largely controlled by the element concentrations in the magma source and by the processes of melt generation and evolution. The world’s largest primary deposits of zirconium and hafnium are associated with alkaline igneous rocks, and, in one locality on the Kola Peninsula of Murmanskaya Oblast, Russia, baddeleyite is recovered as a byproduct of apatite and magnetite mining. Otherwise, there are few primary igneous deposits of zirconium- and hafnium-bearing minerals with economic value at present. The main ore deposits worldwide are heavy-mineral sands produced by the weathering and erosion of preexisting rocks and the concentration of zircon and other economically important heavy minerals, such as ilmenite and rutile (for titanium), chromite (for chromium), and monazite (for rare-earth elements) in sedimentary systems, particularly in coastal environments. In coastal deposits, heavy-mineral enrichment occurs where sediment is repeatedly reworked by wind, waves, currents, and tidal processes. The resulting heavy-mineral-sand deposits, called placers or paleoplacers, preferentially form at relatively low latitudes on passive continental margins and supply 100 percent of

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.

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.

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.

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.

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.

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.

Appling grain boundary engineering to Alloy 690 tube for enhancing intergranular corrosion resistance

NASA Astrophysics Data System (ADS)

Xia, Shuang; Li, Hui; Liu, Ting Guang; Zhou, Bang Xin

2011-09-01

The feasibility of applying the grain boundary engineering (GBE) processing to Alloy 690 tube manufacturing for improving the intergranular corrosion resistance was studied. Through small amount of deformation by cold drawing using a draw-bench on a production line and subsequent short time annealing at high temperature, the proportion of low Σ coincidence site lattice (CSL) grain boundaries of the Alloy 690 tube can be enhanced to about 75% which mainly were of Σ3 n ( n = 1, 2, 3, …) type. In this case, the grain boundary network (GBN) was featured by the formation of highly twinned large size grain-clusters produced by multiple twinning during recrystallization. All of the grains inside this kind of cluster had Σ3 n mutual misorientations, and hence all the boundaries inside the cluster were of Σ3 n type and formed many interconnected Σ3 n type triple junctions. The weight losses due to grain dropping during intergranular corrosion for the samples with the modified GBN were much less than that with conventional microstructure. Based on the characterization by scanning electron microscopy (SEM) and electron backscatter diffraction (EBSD) technique, it was shown that the highly twinned large size grain-cluster microstructure played a key role in enhancing the intergranular corrosion resistance: (1) the large grain-cluster can arrest the penetration of intergranular corrosion; (2) the large grain-cluster can protect the underlying microstructure.

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.

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.

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.

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.

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.

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.

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.

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

A pulsed eddy current probe for inspection of support plates from within Alloy-800 steam generator tubes

NASA Astrophysics Data System (ADS)

Krause, T. W.; Babbar, V. K.; Underhill, P. R.

2014-02-01

Support plate degradation and fouling in nuclear steam generators (SGs) can lead to SG tube corrosion and loss of efficiency. Inspection and monitoring of these conditions can be integrated with preventive maintenance programs, thereby advancing station-life management processes. A prototype pulsed eddy current (PEC) probe, targeting inspection issues associated with SG tubes in SS410 tube support plate structures, has been developed using commercial finite element (FE) software. FE modeling was used to identify appropriate driver and pickup coil configurations for optimum sensitivity to changes in gap and offset for Alloy-800 SG tubes passing through 25 mm thick SS410 support plates. Experimental measurements using a probe that was manufactured based on the modeled configuration, were used to confirm the sensitivity of differential PEC signals to changes in relative position of the tube within the tube support plate holes. Models investigated the effect of shift and tilt of tube with respect to hole centers. Near hole centers and for small shifts, modeled signal amplitudes from the differentially connected coil pairs were observed to change linearly with tube shift. This was in agreement with experimentally measured TEC coil response. The work paves the way for development of a system targeting the inspection and evaluation of support plate structures in steam generators.

Effect of different thermal treatments on the corrosion resistance of alloy 690 tubing

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

Crum, J.R.; Heck, K.A.; Angeliu, T.M.

1990-03-01

A comparison of the carbide precipitation characteristics and corrosion resistance of commercially produced alloy 690 steam generator tubing from various sources, with different thermal treatments, was made. Four thermal treatments within the ranges of 700--720{degree}C (1292--1328{degree}F)/5--5.75 hrs and 871--927{degree}C (1600--1700{degree}F)/10 min were compared to one another and to two as-mill annealed tubes. All tubes were characterized with respect to chemical composition, mechanical properties, and microstructure. Overall carbide precipitation was determined by etching with phosphoric acid/nital, bromine-methanol and glyceregia/oxalic etchants. Scanning transmission electron microscope analysis provided detailed chromium depletion profiles across the grain boundary and carbide composition. Nitric acid intergranular attackmore » (IGA) tests were also conducted. C-ring stress corrosion cracking (SCC) tests, with stresses above the yield strength were then conducted in 350{degree}C (662{degree}F) deaerated 1, 10, and 50% NaOH and unstressed IGA tests were conducted in a NaOH-Na{sub 2}SO{sub 4}--Fe{sub 3}O{sub 4}--Fe{sub 2}O{sub 3} environment, also at 350{degree}C (662{degree}F). 2 tabs.« less

Fiber laser micromachining of magnesium alloy tubes for biocompatible and biodegradable cardiovascular stents

NASA Astrophysics Data System (ADS)

Demir, Ali Gökhan; Previtali, Barbara; Colombo, Daniele; Ge, Qiang; Vedani, Maurizio; Petrini, Lorenza; Wu, Wei; Biffi, Carlo Alberto

2012-02-01

Magnesium alloys constitute an attractive solution for cardiovascular stent applications due to their intrinsic properties of biocompatibility and relatively low corrosion resistance in human-body fluids, which results in as a less intrusive treatment. Laser micromachining is the conventional process used to cut the stent mesh, which plays the key role for the accurate reproduction of the mesh design and the surface quality of the produced stent that are important factors in ensuring the mechanical and corrosion resistance properties of such a kind of devices. Traditionally continuous or pulsed laser systems working in microsecond pulse regime are employed for stent manufacturing. Pulsed fiber lasers on the other hand, are a relatively new solution which could balance productivity and quality aspects with shorter ns pulse durations and pulse energies in the order of mJ. This work reports the study of laser micromachining and of AZ31 magnesium alloy for the manufacturing of cardiovascular stents with a novel mesh design. A pulsed active fiber laser system operating in nanosecond pulse regime was employed for the micromachining. Laser parameters were studied for tubular cutting on a common stent material, AISI 316L tubes with 2 mm in diameter and 0.2 mm in thickness and on AZ31 tubes with 2.5 mm in diameter and 0.2 in thickness. In both cases process parameters conditions were examined for reactive and inert gas cutting solutions and the final stent quality is compared.

Results from Alloy 600 And Alloy 690 Caustic SCC Model Boiler Tests

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

Miller, Frederick D.; Thomas, Larry E.

2009-08-03

A versatile model boiler test methodology was developed and used to compare caustic stress corrosion cracking (SCC) of mill annealed Alloy 600 and thermally treated Alloy 690. The model boiler included simulated crevice devices that efficiently and consistently concentrated Na2CO3, resulting in volatilization of CO2 with the steam and concentration of NaOH at the tube surfaces. The test methodology also included variation in tube stress, either produced by the primary to secondary side pressure differential, or by a novel method that reproducibly yields a higher stress condition on the tube. The significant effect of residual stress on tube SCC wasmore » also considered. SCC of both Alloy 600 and Alloy 690 were evaluated as a function of temperature and stress. Analytical transmission electron microscopy (ATEM) evaluations of the cracks and the grain boundaries ahead of the cracks were performed, providing insight into the SCC mechanism. This model boiler test methodology may be applicable to a range of bulkwater secondary chemistries that concentrate to produce aggressive crevice environments.« less

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.

Strong, corrosion-resistant aluminum tubing

NASA Technical Reports Server (NTRS)

Reed, M. W.; Adams, F. F.

1980-01-01

When aluminum tubing having good corrosion resistance and postweld strength is needed, type 5083 alloy should be considered. Chemical composition is carefully controlled and can be drawn into thin-wall tubing with excellent mechanical properties. Uses of tubing are in aircraft, boats, docks, and process equipment.

Creep-Rupture Behavior and Recrystallization in HR6W and Haynes Alloy 230 Cold-Bent Boiler Tubing for Ultrasupercritical (USC) Steam Boiler Applications

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

Shingledecker, John P

2007-01-01

Creep-rupture experiments were conducted on HR6W and Haynes 230, candidate Ultrasupercritical (USC) alloys, tubes to evaluate the effects of cold-work and recrystallization during high-temperature service. These creep tests were performed by internally pressurizing cold-bent boiler tubes at 775 C for times up to 8000 hours. The bends were fabricated with cold-work levels beyond the current ASME Boiler and Pressure Vessel (ASME B&PV) Code Section I limits for austenitic stainless steels. Destructive metallographic evaluation of the crept tube bends was used to determine the effects of cold-work and the degree of recrystallization. The metallographic analysis combined with an evaluation of themore » creep and rupture data suggest that solid-solution strengthened nickel-based alloys can be fabricated for high-temperature service at USC conditions utilizing levels of cold-work higher than the current allowed levels for austenitic stainless steels.« less

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.

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.

Corrosion of Stainless-Steel Tubing in a Spacecraft Launch Environment

NASA Technical Reports Server (NTRS)

Barile, Ronald G.; MacDowell, Louis G.; Curran, Joe; Calle, Luz Maria; Hodge, Timothy

2001-01-01

This is a report of exposure of various metal tubing to oceanfront launch environments. The objective is to examine various types of corrosion-resistant tubing for Space Shuttle launch sites. The metals were stainless steels (austenitic, low-carbon, Mo-alloy, superaustenitic, duplex, and superferritic), Ni-Cr-Mo alloy, Ni-Mo-Cr-Fe-W alloy, and austenitic Ni-base superalloy.

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.

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.

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.

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.

High temperature alloy

NASA Technical Reports Server (NTRS)

Frank, R. G.; Semmel, J. W., Jr.

1968-01-01

Molybdenum is substituted for tungsten on an atomic basis in a cobalt-based alloy, S-1, thus enabling the alloy to be formed into various mill products, such as tubing and steels. The alloy is weldable, has good high temperature strength and is not subject to embrittlement produced by high temperature aging.

Development and quality assessments of commercial heat production of ATF FeCrAl tubes

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

Yamamoto, Yukinori

2015-09-01

Development and quality assessment of the 2 nd generation ATF FeCrAl tube production with commercial manufacturers were conducted. The manufacturing partners include Sophisticated Alloys, Inc. (SAI), Butler, PA for FeCrAl alloy casting via vacuum induction melting, Oak Ridge National Laboratory (ORNL) for extrusion process to prepare the master bars/tubes to be tube-drawn, and Rhenium Alloys, Inc. (RAI), North Ridgeville, OH, for tube-drawing process. The masters bars have also been provided to Los Alamos National Laboratory (LANL) who works with Century Tubes, Inc., (CTI), San Diego, CA, as parallel tube production effort under the current program.

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.

The stability of alloying additions in Zirconium

NASA Astrophysics Data System (ADS)

Lumley, S. C.; Murphy, S. T.; Burr, P. A.; Grimes, R. W.; Chard-Tuckey, P. R.; Wenman, M. R.

2013-06-01

The interactions of Cr, Fe, Nb, Ni, Sn, V and Y with Zr are simulated using density functional theory. Thermodynamic stabilities of various different Zr based intermetallic compounds, including multiple Laves phase structures and solutions of alloying additions in both α and β-Zr were investigated. The thermodynamic driving forces in this system can be correlated with trends in atomic radii and the relative electronegativities of the different species. Formation energies of Fe, Ni and Sn based intermetallic compounds were found to be negative, and the ZrFe and ZrNi intermetallics were metastable. Most elements displayed negative energies of solution in β-Zr but positive energies in the α-phase, with the exception of Sn (which was negative for both) and Y (which was positive for both). Solutions formed from intermetallics showed a similar trend. Cr -3s23p64s13d5. Fe -4s23d6. Nb -4s24p65s14d4. Ni -4s23d8. Sn -5s25p2. V -3s23p64s23d3. Y -4s24p65s24d1. Zr -4s24p65s24d2. The pseudopotential scheme used is "on-the-fly" generation, in which an isolated all-electron calculation is carried out before the main calculation and used as a starting point to generate a pseudopotential. This was carried out for all pseudopotentials except Cr and V, as the default on-the-fly pseudopotentials for these elements required a much higher cut-off energy. Instead, standard ultrasoft pseudopotentials, as found in the CASTEP pseudopotential library, were used for Cr and V. All pseudopotentials (both on-the-fly and library) are of the ultrasoft type [15], and so are compatible with each-other. Exchange-correlation was modelled using the Perdew, Burke and Ernzerhof formalisation of the Generalised Gradient Approximation [16].A series of simulations were run to establish an appropriate basis set cut-off energy, and the density of sampling in the Brillouin zone. The results were converged to within two decimal places for a cut-off energy of 450 eV and a k-point spacing of 0.003 nm-1. The k

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.

Evolution of grain boundary character distributions in alloy 825 tubes during high temperature annealing: Is grain boundary engineering achieved through recrystallization or grain growth?

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

Bai, Qin; Zhao, Qing

Grain boundary engineering (GBE) of nickel-based alloy 825 tubes was carried out with different cold drawing deformations by using a draw-bench on a factory production line and subsequent annealing at various temperatures. The microstructure evolution of alloy 825 during thermal-mechanical processing (TMP) was characterized by means of the electron backscatter diffraction (EBSD) technique to study the TMP effects on the grain boundary network and the evolution of grain boundary character distributions during high temperature annealing. The results showed that the proportion of ∑ 3{sup n} coincidence site lattice (CSL) boundaries of alloy 825 tubes could be increased to > 75%more » by the TMP of 5% cold drawing and subsequent annealing at 1050 °C for 10 min. The microstructures of the partially recrystallized samples and the fully recrystallized samples suggested that the proportion of low ∑ CSL grain boundaries depended on the annealing time. The frequency of low ∑ CSL grain boundaries increases rapidly with increasing annealing time associating with the formation of large-size highly-twinned grains-cluster microstructure during recrystallization. However, upon further increasing annealing time, the frequency of low ∑ CSL grain boundaries decreased markedly during grain growth. So it is concluded that grain boundary engineering is achieved through recrystallization rather than grain growth. - Highlights: •The grain boundary engineering (GBE) is applicable to 825 tubes. •GBE is achieved through recrystallization rather than grain growth. •The low ∑ CSL grain boundaries in 825 tubes can be increased to > 75%.« less

Method of forming magnetostrictive rods from rare earth-iron alloys

DOEpatents

McMasters, O.D.

1986-09-02

Rods of magnetostrictive alloys of iron with rare earth elements are formed by flowing a body of rare earth-iron alloy in a crucible enclosed in a chamber maintained under an inert gas atmosphere, forcing such molten rare-earth-iron alloy into a hollow mold tube of refractory material positioned with its lower end portion within the molten body by means of a pressure differential between the chamber and mold tube and maintaining a portion of the molten alloy in the crucible extending to a level above the lower end of the mold tube so that solid particles of higher melting impurities present in the alloy collect at the surface of the molten body and remain within the crucible as the rod is formed in the mold tube. 5 figs.

Method of forming magnetostrictive rods from rare earth-iron alloys

DOEpatents

McMasters, O. Dale

1986-09-02

Rods of magnetrostructive alloys of iron with rare earth elements are formed by flowing a body of rare earth-iron alloy in a crucible enclosed in a chamber maintained under an inert gas atmosphere, forcing such molten rare-earth-iron alloy into a hollow mold tube of refractory material positioned with its lower end portion within the molten body by means of a pressure differential between the chamber and mold tube and maintaining a portion of the molten alloy in the crucible extending to a level above the lower end of the mold tube so that solid particles of higher melting impurities present in the alloy collect at the surface of the molten body and remain within the crucible as the rod is formed in the mold tube.

A Microstructure Study on an AZ31 Magnesium Alloy Tube after Hot Metal Gas Forming Process

NASA Astrophysics Data System (ADS)

Liu, Yi; Wu, Xin

2007-06-01

An AZ31 magnesium alloy tube has been deformed by the hot metal gas forming (HMGF) technique. Microstructures before and after deformation have been investigated by using Electron Backscattered Diffraction (EBSD) and Electron Microscopy. Due to the inhomogeneous distribution by induction heating, there is a temperature gradient distribution along the tube axis. Accordingly, the deformation mechanism is also different. In the middle area of deformation zone where the temperature is ˜410 °C, almost no twinning has been found, whereas at the edge areas of deformation zone where the temperature is ˜200 °C, a high density of twins has been found. EBSD experiments show a weak (0001) fiber texture along the radial direction of the tube before and after deformation in the high-temperature zone. EBSD experiments on the low temperature deformation region were not successful due to the high stored energy. Schmid factor analysis on the EBSD data shows that, despite the (0001) fiber texture, there are still many grains favoring basal slip along both the axis direction and hoop direction.

Substantially oxygen-free contact tube

NASA Technical Reports Server (NTRS)

Pike, James F. (Inventor)

1993-01-01

A device for arc welding is provided in which a continuously-fed electrode wire is in electrical contact with a contact tube. The contact tube is improved by using a substantially oxygen-free conductive alloy in order to reduce the amount of electrical erosion.

Substantially Oxygen-Free Contact Tube

NASA Technical Reports Server (NTRS)

Pike, James F. (Inventor)

1991-01-01

A device for arc welding is provided in which a continuously-fed electrode wire is in electrical contact with a contact tube. The contact tube is improved by using a substantially oxygen-free conductive alloy in order to reduce the amount of electrical erosion.

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

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.

Torsion Tests of Tubes

NASA Technical Reports Server (NTRS)

Stang, Ambrose H; Ramberg, Walter; Back, Goldie

1937-01-01

This report presents the results of tests of 63 chromium-molybdenum steel tubes and 102 17st aluminum-alloy tubes of various sizes and lengths made to study the dependence of the torsional strength on both the dimensions of the tube and the physical properties of the tube material. Three types of failure are found to be important for sizes of tubes frequently used in aircraft construction: (1) failure by plastic shear, in which the tube material reached its yield strength before the critical torque was reached; (2) failure by elastic two-lobe buckling, which depended only on the elastic properties of the tube material and the dimensions of the tube; and (3) failure by a combination of (1) and (2) that is, by buckling taking place after some yielding of the tube material.

Hydrostatic Extrusion of 60mm Mortar Tubes

DTIC Science & Technology

1974-10-01

Transverse "Tensile Test Specimen 16 9. Subsize Charpy and Tensile Test Specimens 17 10. Extruded 718 Alloy Tube Pressure-strain Pressure Test [)ata 18... subsize specimens. (d) Previous studies have shown that the standard size Charpy values are 4.5 times the subsize values. 22 -. •. ., *. ...20 II. Tensile Properties of Extruded and Aged Tube (a) 21 III. Extruded 718 Alloy Pressure Test Data 21 IV. V-Notch Charpy Impact Properties(c) of

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.

Leakproof Swaged Joints in Thin-Wall Tubing

NASA Technical Reports Server (NTRS)

Stuckenberg, F. H.; Crockett, L. K.; Snyder, W. E.

1986-01-01

Tubular inserts reinforce joints, reducing incidence of leaks. In new swaging technique, tubular inserts placed inside ends of both tubes to be joined. Made from thicker-wall tubing with outside diameter that matches inside diameter of thin tubing swaged, inserts support tube ends at joint. They ensure more uniform contact between swage fitting and tubing. New swaging technique developed for Al/Ti/V-alloy hydraulic supply lines.

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.

Role of lead in electrochemical reaction of alloy 600, alloy 690, Ni, Cr, and Fe in water

NASA Astrophysics Data System (ADS)

Hwang, Seong Sik; Kim, Joung Soo; Kim, Ju Yup

2003-08-01

It has been reported that lead causes stress corrosion cracking (SCC) in the secondary side of steam generators (SG) in pressurized water reactors (PWR). The materials of SG tubings are alloy 600, alloy 690, or alloy 800, among which the main alloying elements are Ni, Cr, and Fe. The effect of lead on the electrochemical behaviors of alloy 600 and alloy 690 using an anodic polarization technique was evaluated. We also obtained polarization curves of pure Ni, Cr, and Fe in water containing lead. As the amount of lead in the solution increased, critical current densities and passive current densities of alloy 600 and alloy 690 increased, while the breakdown potential of the alloys decreased. Lead increased critical current density and the passive current of Cr in pH 4 and pH 10. The instability of passive film of steam generator tubings in water containing lead might arise from the instability of Cr passivity.

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.

Research on Protective Coating on Inner Surface of Alloy Tube

NASA Astrophysics Data System (ADS)

Zhang, Y. C.; Liu, Y. H.; Zhou, Z. J.; Zheng, M. M.; Kong, S. Y.; Xia, H. H.; Li, H. L.

2017-09-01

Materials are one of the most important factors which limit reactor development. Molten salt not only used as the coolant but used as application in which fissile materials and fission products are dissolved in Molten Salt Reactors (MSRs). Therefore the corrosion resistance of structure materials is the one of most important aspects for application in MSRs. Compatibility and chemical stability with the molten salt should be considered for some common structural alloys such as Incoloy-800H. In this research, the pure nickel coating was obtained by electroplating on the inner surface of nickel alloy to improve the corrosion resistance. However, there are some problems for plating on the inner surface of tube. For example the current is shielded and the anode is easy to passivate. The inner anode was used for solving these problems in this study. Pure nickel coating was obtain and the microstructure and properties of coating were analysed using this method. The thickness, hardness and microstructure of coating were observed by metallographic microscope, micro hardness tester and field emission scanning electron microscope, and the influence of deposition duration and annealing treatment duration on properties were analysed. Thermal shock performance was investigated as well. The results showed that the coating thickness increased linearly with the increasing of plating durations and the size of grain increased with the durations as well, the surface of coating became inhomogeneous correspondingly. The hardness of coating changed as the change of durations of annealing treatment. The thermal shock test showed that bonding strength of coating with substrate was good.

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.

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.

Measurement of fracture stress for 6000-series extruded aluminum alloy tube using multiaxial tube expansion testing method

NASA Astrophysics Data System (ADS)

Nagai, Keisuke; Kuwabara, Toshihiko; Ilinich, Andrey; Luckey, George

2018-05-01

A servo-controlled tension-internal pressure testing machine with an optical 3D digital image correlation system (DIC) is used to measure the multiaxial deformation behavior of an extruded aluminum alloy tube for a strain range from initial yield to fracture. The outer diameter of the test sample is 50.8 mm and wall thickness 2.8 mm. Nine linear stress paths are applied to the specimens: σɸ (axial true stress component) : σθ (circumferential true stress component) = 1:0, 4:1, 2:1, 4:3, 1:1, 3:4, 1:2, 1:4, and 0:1. The equivalent strain rate is approximately 5 × 10-4 s-1 constant. The forming limit curve (FLC) and forming limit stress curve (FLSC) are also measured. Moreover, the average true stress components inside a localized necking area are determined for each specimen from the thickness strain data for the localized necking area and the geometry of the fracture surface.

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.

Methods for batch fabrication of cold cathode vacuum switch tubes

DOEpatents

Walker, Charles A [Albuquerque, NM; Trowbridge, Frank R [Albuquerque, NM

2011-05-10

Methods are disclosed for batch fabrication of vacuum switch tubes that reduce manufacturing costs and improve tube to tube uniformity. The disclosed methods comprise creating a stacked assembly of layers containing a plurality of adjacently spaced switch tube sub-assemblies aligned and registered through common layers. The layers include trigger electrode layer, cathode layer including a metallic support/contact with graphite cathode inserts, trigger probe sub-assembly layer, ceramic (e.g. tube body) insulator layer, and metallic anode sub-assembly layer. Braze alloy layers are incorporated into the stacked assembly of layers, and can include active metal braze alloys or direct braze alloys, to eliminate costs associated with traditional metallization of the ceramic insulator layers. The entire stacked assembly is then heated to braze/join/bond the stack-up into a cohesive body, after which individual switch tubes are singulated by methods such as sawing. The inventive methods provide for simultaneously fabricating a plurality of devices as opposed to traditional methods that rely on skilled craftsman to essentially hand build individual devices.

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.

Fine-grained zirconium-base material

DOEpatents

Van Houten, G.R.

1974-01-01

A method is described for making zirconium with inhibited grain growth characteristics, by the process of vacuum melting the zirconium, adding 0.3 to 0.5% carbon, stirring, homogenizing, and cooling. (Official Gazette)

Heat removal capability of divertor coaxial tube assembly

NASA Astrophysics Data System (ADS)

Shibui, Masanao; Nakahira, Masataka; Tada, Eisuke; Takatsu, Hideyuki

1994-05-01

To deal with high power flowing in the divertor region, an advanced divertor concept with gas target has been proposed for use in ITER/EDA. The concept uses a divertor channel to remove the radiated power while allowing neutrals to recirculate. Candidate channel wall designs include a tube array design where many coaxial tubes are arranged in the toroidal direction to make louver. The coaxial tube consists of a Be protection tube encases many supply tubes wound helically around a return tube. V-alloy and hardened Cu-alloy have been proposed for use in the supply and return tubes. Some coolants have also been proposed for the design including pressurized He and liquid metals, because these coolants are consistent with the selection of coolants for the blanket and also meet the requirement of high temperature operation. In the coaxial tube design, the coolant area is restricted and brittle Be material is used under severe thermal cyclings. Thus, to obtain the coaxial tube with sufficient safety margin for the expected fusion power excursion, it is essential to understand its applicability limit. The paper discusses heat removal capability of the coaxial tube and recommends some design modifications.

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

SEPARATION OF HAFNIUM FROM ZIRCONIUM

DOEpatents

Overholser, L.B.; Barton, C.J. Sr.; Ramsey, J.W.

1960-05-31

The separation of hafnium impurities from zirconium can be accomplished by means of organic solvent extraction. The hafnium-containing zirconium feed material is dissolved in an aqueous chloride solution and the resulting solution is contacted with an organic hexone phase, with at least one of the phases containing thiocyanate. The hafnium is extracted into the organic phase while zirconium remains in the aqueous phase. Further recovery of zirconium is effected by stripping the onganic phase with a hydrochloric acid solution and commingling the resulting strip solution with the aqueous feed solution. Hexone is recovered and recycled by means of scrubbing the onganic phase with a sulfuric acid solution to remove the hafnium, and thiocyanate is recovered and recycled by means of neutralizing the effluent streams to obtain ammonium thiocyanate.

Zirconium

USGS Publications Warehouse

Bedinger, G.M.

2013-01-01

Zirconium is the 20th most abundant element in the Earth’s crust. It occurs in a variety of rock types and geologic environments but most often in igneous rocks in the form of zircon (ZrSiO4). Zircon is recovered as a coproduct of the mining and processing of heavy mineral sands for the titanium minerals ilmenite and rutile. The sands are formed by the weathering and erosion of rock containing zircon and titanium heavy minerals and their subsequent concentration in sedimentary systems, particularly in coastal environments. A small quantity of zirconium, less than 10 kt/a (11,000 stpy), compared with total world production of 1.4 Mt (1.5 million st) in 2012, was derived from the mineral baddeleyite (ZrO2), produced from a single source in Kovdor, Russia.

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

IGA resistance of TT Alloy 690 and concentration behavior of Broached Egg Crate tube support configuration

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

Suzuki, S.; Kusakabe, T.; Yamamoto, H.

1992-12-31

In order to improve the reliability of the Steam Generator (SG), TT Alloy 690 and BEC (Broached Egg Crate) type tube support plate has been developed. Some tests are carried out to heighten the reliability for these improvements all the more and the following results are obtained. (1) SERT test (Slow Extension Rate Test) made clear that TT690 has less IGA susceptibility in comparison with MA600. (2) The alkaline susceptibility on the occurrence of IGA/SCC on TT690 and MA600 obtained by SERT corresponds to that obtained by Model Boiler test. (3) By model boiler test, superior concentration behaviors for BECmore » type tube support plate configuration have been recognized in comparison with Drill type. This result is obtained by the joint research of the five utilities (Kansai Epco, Hokkaido Epco, Shikoku Epco, Kyushu Epco, JAPCO) and MHI.« less

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.

Alloy 690 qualification: Corrosion under prototypic heat flux and temperature conditions. Final report

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

Baum, A.J.

1995-05-01

The objective of this program was to perform qualification tests of industrial heats of thermally treated alloy 690 steam generator tubing under heat transfer conditions. Primary emphasis was focused on testing of alternate tube to tubesheet expansion processes. In addition, a background report was written to document the evolution of the alloy 690 process specification and the supporting qualification testing. While the testing was able to produce some localized corrosion of alloy 690 in the tube to tubesheet joint transition regions under highly alkaline conditions, the corrosion rates were between two and three orders of magnitude lower than the comparablemore » rates in mill annealed alloy 600 tubing. The corrosion morphology was a combination of intergranular and general corrosion, rather than the stress corrosion cracking typically found in mill annealed alloy 600 tubing.« less

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...

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.

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.

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.

Acoustic emission analysis on tensile failure of steam-side oxide scales formed on T22 alloy superheater tubes

NASA Astrophysics Data System (ADS)

Huang, Jun-Lin; Zhou, Ke-Yi; Wang, Xin-Meng; Tu, Yi-You; Xu, Jian-Qun

2014-07-01

Failure of steam-side oxide scales on boiler tubes can seriously influence the safety of coal-fired power plants. Uniaxial tensile tests employing acoustic emission (AE) monitoring were performed, in this work, to investigate the failure behavior of steam-side oxide scales on T22 alloy boiler superheater tubes. The characteristic frequency spectra of the captured AE signals were obtained by performing fast Fourier transform. Three distinct peak frequency bands, 100-170, 175-250, and 280-390 kHz, encountered in different testing stages were identified in the frequency spectra, which were confirmed to, respectively, correspond to substrate plastic deformation, oxide vertical cracking, and oxide spalling with the aid of scanning electronic microscopy observations, and can thus be used for distinguishing different oxide failure mechanisms. Finally, the critical cracking strain of the oxide scale and the interfacial shear strength of the oxide/substrate interface were estimated, which are the critical parameters urgently desired for modeling the failure behavior of steam-side oxide scales on boiler tubes of coal-fired power plants.

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%.

Technique development for modulus, microcracking, hermeticity, and coating evaluation capability characterization of SiC/SiC tubes

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

Hu, Xunxiang; Ang, Caen K.; Singh, Gyanender P.

Driven by the need to enlarge the safety margins of nuclear fission reactors in accident scenarios, research and development of accident-tolerant fuel has become an important topic in the nuclear engineering and materials community. A continuous-fiber SiC/SiC composite is under consideration as a replacement for traditional zirconium alloy cladding owing to its high-temperature stability, chemical inertness, and exceptional irradiation resistance. An important task is the development of characterization techniques for SiC/SiC cladding, since traditional work using rectangular bars or disks cannot directly provide useful information on the properties of SiC/SiC composite tubes for fuel cladding applications. At Oak Ridge Nationalmore » Laboratory, experimental capabilities are under development to characterize the modulus, microcracking, and hermeticity of as-fabricated, as-irradiated SiC/SiC composite tubes. Resonant ultrasound spectroscopy has been validated as a promising technique to evaluate the elastic properties of SiC/SiC composite tubes and microcracking within the material. A similar technique, impulse excitation, is efficient in determining the basic mechanical properties of SiC bars prepared by chemical vapor deposition; it also has potential for application in studying the mechanical properties of SiC/SiC composite tubes. Complete evaluation of the quality of the developed coatings, a major mitigation strategy against gas permeation and hydrothermal corrosion, requires the deployment of various experimental techniques, such as scratch indentation, tensile pulling-off tests, and scanning electron microscopy. In addition, a comprehensive permeation test station is being established to assess the hermeticity of SiC/SiC composite tubes and to determine the H/D/He permeability of SiC/SiC composites. This report summarizes the current status of the development of these experimental capabilities.« less

Oxidized zirconium on ceramic; Catastrophic coupling.

PubMed

Ozden, V E; Saglam, N; Dikmen, G; Tozun, I R

2017-02-01

Oxidized zirconium (Oxinium™; Smith & Nephew, Memphis, TN, USA) articulated with polyethylene in total hip arthroplasty (THA) appeared to have the potential to reduce wear dramatically. The thermally oxidized metal zirconium surface is transformed into ceramic-like hard surface that is resistant to abrasion. The exposure of soft zirconium metal under hard coverage surface after the damage of oxidized zirconium femoral head has been described. It occurred following joint dislocation or in situ succeeding disengagement of polyethylene liner. We reported three cases of misuse of Oxinium™ (Smith & Nephew, Memphis, TN, USA) heads. These three cases resulted in catastrophic in situ wear and inevitable failure although there was no advice, indication or recommendation for this use from the manufacturer. Copyright © 2016 Elsevier Masson SAS. All rights reserved.

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.

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.

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.

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

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.

Separation of Zirconium and Hafnium: A Review

NASA Astrophysics Data System (ADS)

Xu, L.; Xiao, Y.; van Sandwijk, A.; Xu, Q.; Yang, Y.

Zirconium is an ideal material for nuclear reactors due to its low absorption cross-section for thermal neutrons, whereas the typically contained hafnium with strong neutron-absorption is very harmful for zirconium. This paper provides an overview of the processes for separating hafnium from zirconium. The separation processes are roughly classified into hydro- and pyrometallurgical routes. The current dominant zirconium production route involves pyrometallurgical ore cracking, multi-step hydrometallurgical liquid-liquid extraction for hafnium removal and the reduction of zirconium tetrachloride to the pure metal by the Kroll process. The lengthy hydrometallurgical Zr-Hf separation operations leads to high production cost, intensive labour and heavy environmental burden. Using a compact pyrometallurgical separation method can simplify the whole production flowsheet with a higher process efficiency. The known separation methods are discussed based on the following reaction features: redox characteristics, volatility, electrochemical properties and molten salt extraction. The commercially operating extractive distillation process is a significant advance in Zr-Hf separation technology but it suffers from high process maintenance cost. The recently developed new process based on molten salt-metal equilibrium for Zr-Hf separation shows a great potential for industrial application, which is compact for nuclear grade zirconium production starting from crude ore. In the present paper, the available separation technologies are compared. The advantages and disadvantages as well as future directions of research and development for nuclear grade zirconium production are discussed.

SEPARATING HAFNIUM FROM ZIRCONIUM

DOEpatents

Lister, B.A.J.; Duncan, J.F.

1956-08-21

A dilute aqueous solution of zirconyl chloride which is 1N to 2N in HCl is passed through a column of a cation exchange resin in acid form thereby absorbing both zirconium and associated hafnium impurity in the mesin. The cation exchange material with the absorbate is then eluted with aqueous sulfuric acid of a O.8N to 1.2N strength. The first portion of the eluate contains the zirconium substantially free of hafnium.

[A surface reacted layer study of titanium-zirconium alloy after dental casting].

PubMed

Zhang, Y; Guo, T; Li, Z; Li, C

2000-10-01

To investigate the influence of the mold temperature on the surface reacted layer of Ti-Zr alloy castings. Ti-Zr alloy was casted into a mold which was made of a zircon (ZrO2.SiO2) for inner coating and a phosphate-bonded material for outer investing with a casting machine (China) designed as vacuum, pressure and centrifuge. At three mold temperatures (room temperature, 300 degrees C, 600 degrees C) the Ti-Zr alloy was casted separately. The surface roughness of the castings was calculated by instrument of smooth finish (China). From the surface to the inner part the Knoop hardness and thickness in reacted layer of Ti-Zr alloy casting was measured. The structure of the surface reacted layer was analysed by SEM. Elemental analyses of the interfacial zone of the casting was made by element line scanning observation. The surface roughness of the castings was increased significantly with the mold temperature increasing. At a higher mold temperature the Knoop hardness of the reactive layer was increased. At the three mold temperature the outmost surface was very hard, and microhardness data decreased rapidly where they reached constant values. The thickness was about 85 microns for castings at room temperature and 300 degrees C, 105 microns for castings at 600 degrees C. From the SEM micrograph of the Ti-Zr alloy casting, the surface reacted layer could be divided into three different layers. The first layer was called non-structure layer, which thickness was about 10 microns for room temperature group, 20 microns for 300 degrees C and 25 microns for 600 degrees C. The second layer was characterized by coarse-grained acicular crystal, which thickness was about 50 microns for three mold temperatures. The third layer was Ti-Zr alloy. The element line scanning showed non-structure layer with higher level of element of O, Al, Si and Zr, The higher the mold temperature during casting, the deeper the Si permeating and in the second layer the element Si could also be found

Physical and mechanical metallurgy of zirconium alloys for nuclear applications: a multi-scale computational study

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

Glazoff, Michael Vasily

2014-10-01

In the post-Fukushima world, the stability of materials under extreme conditions is an important issue for the safety of nuclear reactors. Because the nuclear industry is going to continue using advanced zirconium cladding materials in the foreseeable future, it become critical to gain fundamental understanding of the several interconnected problems. First, what are the thermodynamic and kinetic factors affecting the oxidation and hydrogen pick-up by these materials at normal, off-normal conditions, and in long-term storage? Secondly, what protective coatings (if any) could be used in order to gain extremely valuable time at off-normal conditions, e.g., when temperature exceeds the criticalmore » value of 2200°F? Thirdly, the kinetics of oxidation of such protective coating or braiding needs to be quantified. Lastly, even if some degree of success is achieved along this path, it is absolutely critical to have automated inspection algorithms allowing identifying defects of cladding as soon as possible. This work strives to explore these interconnected factors from the most advanced computational perspective, utilizing such modern techniques as first-principles atomistic simulations, computational thermodynamics of materials, diffusion modeling, and the morphological algorithms of image processing for defect identification. Consequently, it consists of the four parts dealing with these four problem areas preceded by the introduction and formulation of the studied problems. In the 1st part an effort was made to employ computational thermodynamics and ab initio calculations to shed light upon the different stages of oxidation of ziraloys (2 and 4), the role of microstructure optimization in increasing their thermal stability, and the process of hydrogen pick-up, both in normal working conditions and in long-term storage. The 2nd part deals with the need to understand the influence and respective roles of the two different plasticity mechanisms in Zr nuclear alloys

Effect of tube processing methods on the texture and grain boundary characteristics of 14YWT nanostructured ferritic alloys

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

Aydogan, E.; Pal, S.; Anderoglu, O.

In this paper, texture and microstructure of tubes and plates fabricated from a nanostructured ferritic alloy (14YWT), produced either by spray forming followed by hydrostatic extrusion (Process I) or hot extrusion and cross-rolling a plate followed by hydrostatic tube extrusion (Process II) have been characterized in terms of their effects on texture and grain boundary character. Hydrostatic extrusion results in a combination of plane strain and shear deformations which generate low intensity α- and γ-fiber components of {001}<110> and {111}<110> together with a weak ζ-fiber component of {011}<211> and {011}<011>. In contrast, multi-step plane strain deformation by hot extrusion andmore » cross-rolling of the plate leads to a strong texture component of {001}<110> together with a weaker {111}<112> component. Although the total strains are similar, shear dominated deformation leads to much lower texture indexes compared to plane strain deformations. Further, the texture intensity decreases after hydrostatic extrusion of the alloy plate formed by plane strain deformation, due to a lower number of activated slip systems during shear dominated deformation. Finally and notably, hot extruded and cross-rolled plate subjected to plane strain deformation to ~50% engineering strain creates only a modest population of low angle grain boundaries, compared to the much larger population observed following the combination of plane strain and shear deformation of ~44% engineering strain resulting from subsequent hydrostatic extrusion.« less

Effect of tube processing methods on the texture and grain boundary characteristics of 14YWT nanostructured ferritic alloys

DOE PAGES

Aydogan, E.; Pal, S.; Anderoglu, O.; ...

2016-03-08

In this paper, texture and microstructure of tubes and plates fabricated from a nanostructured ferritic alloy (14YWT), produced either by spray forming followed by hydrostatic extrusion (Process I) or hot extrusion and cross-rolling a plate followed by hydrostatic tube extrusion (Process II) have been characterized in terms of their effects on texture and grain boundary character. Hydrostatic extrusion results in a combination of plane strain and shear deformations which generate low intensity α- and γ-fiber components of {001}<110> and {111}<110> together with a weak ζ-fiber component of {011}<211> and {011}<011>. In contrast, multi-step plane strain deformation by hot extrusion andmore » cross-rolling of the plate leads to a strong texture component of {001}<110> together with a weaker {111}<112> component. Although the total strains are similar, shear dominated deformation leads to much lower texture indexes compared to plane strain deformations. Further, the texture intensity decreases after hydrostatic extrusion of the alloy plate formed by plane strain deformation, due to a lower number of activated slip systems during shear dominated deformation. Finally and notably, hot extruded and cross-rolled plate subjected to plane strain deformation to ~50% engineering strain creates only a modest population of low angle grain boundaries, compared to the much larger population observed following the combination of plane strain and shear deformation of ~44% engineering strain resulting from subsequent hydrostatic extrusion.« less

Complete Status Report Documenting Development of Friction Stir Welding for Joining Thin Wall Tubing of ODS Alloys

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

Hoelzer, David T.; Bunn, Jeffrey R.; Gussev, Maxim N.

The development of friction stir welding (FSW) for joining thin sections of the advanced oxide dispersion strengthened (ODS) 14YWT ferritic alloy was initiated in Fuel Cycle Research and Development (FCRD), now the Nuclear Technology Research and Development (NTRD), in 2015. The first FSW experiment was conducted in late FY15 and successfully produced a bead-on-plate stir zone (SZ) on a 1 mm thick plate of 14YWT (SM13 heat). The goal of this research task is to ultimately demonstrate that FSW is a feasible method for joining thin wall (0.5 mm thick) tubing of 14YWT.

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

URANIUM DECONTAMINATION WITH RESPECT TO ZIRCONIUM

DOEpatents

Vogler, S.; Beederman, M.

1961-05-01

A process is given for separating uranium values from a nitric acid aqueous solution containing uranyl values, zirconium values and tetravalent plutonium values. The process comprises contacting said solution with a substantially water-immiscible liquid organic solvent containing alkyl phosphate, separating an organic extract phase containing the uranium, zirconium, and tetravalent plutonium values from an aqueous raffinate, contacting said organic extract phase with an aqueous solution 2M to 7M in nitric acid and also containing an oxalate ion-containing substance, and separating a uranium- containing organic raffinate from aqueous zirconium- and plutonium-containing extract phase.

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.

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.

IMPROVEMENT OF THE EXTRACTION SEPARATION OF URANIUM AND ZIRCONIUM USING ZIRCONIUM-MASKING REAGENTS (in German)

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

Kyrs, M.; Caletka, R.; Selucky, P.

1963-12-01

The masking capacities of a series of reagents were studied in the zirconium extraction with tributyl phosphate solution in the presence of nitric acid. It was established that with many reagents an improvement of the separation of uranium from zirconium could be obtained. The efficiency of the reagents increases in the series tannin, oxalic acid, tiron, pyrogallol, and Arsenazo I. (tr-auth)

Production of nuclear grade zirconium: A review

NASA Astrophysics Data System (ADS)

Xu, L.; Xiao, Y.; van Sandwijk, A.; Xu, Q.; Yang, Y.

2015-11-01

Zirconium is an ideal material for nuclear reactors due to its low absorption cross-section for thermal neutrons, whereas the typically contained hafnium with strong neutron-absorption is very harmful for zirconium as a fuel cladding material. This paper provides an overview of the processes for nuclear grade zirconium production with emphasis on the methods of Zr-Hf separation. The separation processes are roughly classified into hydro- and pyrometallurgical routes. The known pyrometallurgical Zr-Hf separation methods are discussed based on the following reaction features: redox characteristics, volatility, electrochemical properties and molten salt-metal equilibrium. In the present paper, the available Zr-Hf separation technologies are compared. The advantages and disadvantages as well as future directions of research and development for nuclear grade zirconium production are discussed.

Experimental observations on uniaxial whole-life transformation ratchetting and low-cycle stress fatigue of super-elastic NiTi shape memory alloy micro-tubes

NASA Astrophysics Data System (ADS)

Song, Di; Kang, Guozheng; Kan, Qianhua; Yu, Chao; Zhang, Chuanzeng

2015-07-01

In this work, the low-cycle fatigue failure of super-elastic NiTi shape memory alloy micro-tubes with a wall thickness of 150 μm is investigated by uniaxial stress-controlled cyclic tests at human body temperature 310 K. The effects of mean stress, peak stress, and stress amplitude on the uniaxial whole-life transformation ratchetting and fatigue failure of the NiTi alloy are observed. It is concluded that the fatigue life depends significantly on the stress levels, and the extent of martensite transformation and its reverse play an important role in determining the fatigue life. High peak stress or complete martensite transformation shortens the fatigue life.

Effect of Elemental Sulfur and Sulfide on the Corrosion Behavior of Cr-Mo Low Alloy Steel for Tubing and Tubular Components in Oil and Gas Industry.

PubMed

Khaksar, Ladan; Shirokoff, John

2017-04-20

The chemical degradation of alloy components in sulfur-containing environments is a major concern in oil and gas production. This paper discusses the effect of elemental sulfur and its simplest anion, sulfide, on the corrosion of Cr-Mo alloy steel at pH 2 and 5 during 10, 20 and 30 h immersion in two different solutions. 4130 Cr-Mo alloy steel is widely used as tubing and tubular components in sour services. According to the previous research in aqueous conditions, contact of solid sulfur with alloy steel can initiate catastrophic corrosion problems. The corrosion behavior was monitored by the potentiodynamic polarization technique during the experiments. Energy dispersive X-ray spectroscopy (EDS) and scanning electron microscopy (SEM) have been applied to characterize the corrosion product layers after each experiment. The results show that under the same experimental conditions, the corrosion resistance of Cr-Mo alloy in the presence of elemental sulfur is significantly lower than its resistance in the presence of sulfide ions.

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).

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.

The role of hydrogen in zirconium alloy corrosion

NASA Astrophysics Data System (ADS)

Ensor, B.; Lucente, A. M.; Frederick, M. J.; Sutliff, J.; Motta, A. T.

2017-12-01

Hydrogen enters zirconium metal as a result of the corrosion process and forms hydrides when present in quantities above the solubility limit at a given temperature. Zircaloy-4 coupons of different thicknesses (0.4 mm-2.3 mm) but identical chemistry and processing were corroded in autoclave at 360 °C for various times up to 2800 days. Coupons were periodically removed and weighed to determine weight gain, which allows follow of the corrosion kinetics. Coupon thickness differences resulted in different volumetric concentrations of hydrogen, as quantified using hot vacuum extraction. The thinnest coupons, having the highest concentration of hydrogen, demonstrated acceleration in their corrosion kinetics and shorter transition times when compared to thicker coupons. Furthermore, it was seen that the post-transition corrosion rate was increased with increasing hydrogen concentration. Corrosion rates increased only after the terminal solid solubility (TSS) was exceeded for hydrogen in Zircaloy-4 at 360 °C. Therefore, it is hypothesized that the corrosion acceleration is caused by the formation of hydrides. Scanning electron microscope (SEM) examinations of fractured oxide layers demonstrate the oxide morphology changed with hydrogen content, with more equiaxed oxide grains in the high hydrogen samples than in those with lower hydrogen content. Additionally, locations of advanced oxide growth were correlated with locations of hydrides in the metal. A hypothesis is proposed to explain the accelerated corrosion due to the presence of the hydrides, namely that the metal, locally, is less able to accommodate oxide growth stresses and this leads to earlier loss of oxide protectiveness in the form of more frequent oxide kinetic transitions.

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.

Quercetin as colorimetric reagent for determination of zirconium

USGS Publications Warehouse

Grimaldi, F.S.; White, C.E.

1953-01-01

Methods described in the literature for the determination of zirconium are generally designed for relatively large amounts of this element. A good procedure using colorimetric reagent for the determination of trace amounts is desirable. Quercetin has been found to yield a sensitive color reaction with zirconium suitable for the determination of from 0.1 to 50?? of zirconium dioxide. The procedure developed involves the separation of zirconium from interfering elements by precipitation with p-dimethylaminoazophenylarsonic acid prior to its estimation with quercetin. The quercetin reaction is carried out in 0.5N hydrochloric acid solution. Under the operating conditions it is indicated that quercetin forms a 2 to 1 complex with zirconium; however, a 2 to 1 and a 1 to 1 complex can coexist under special conditions. Approximate values for the equilibrium constants of the complexes are K1 = 0.33 ?? 10-5 and K2 = 1.3 ?? 10-9. Seven Bureau of Standards samples of glass sands and refractories were analyzed with excellent results. The method described should find considerable application in the analysis of minerals and other materials for macro as well as micro amounts of zirconium.

Microcrack healing in non-ferrous metal tubes through eddy current pulse treatment.

PubMed

Xu, Wenchen; Yang, Chuan; Yu, Haiping; Jin, Xueze; Guo, Bin; Shan, Debin

2018-04-16

This study proposed a novel method to heal microcrack within Mg alloy tubes using high density eddy current pulse treatment (ECPT). Through electromagnetic induction inside a copper coil connected with a high density pulse power source supply, the high density (greater than 5 × 10 9  A/m 2 ) and short duration eddy current was generated in tube specimens of Mg alloy. The results show that the microcracks in tube specimens was healed evidently and the mechanical properties of the tubes subjected to ECPT were improved simultaneously. The crack healing during ECPT was ascribed to not only the thermal stress around the microcrack tips and the softening or melting of metals in the vicinity of microcrack tips, but also the squeezing action acted by the Lorentz force. In the inward-discharging scheme, both the compressive radial stress and tangential stress induced by the Lorentz force contributed to more sufficient crack healing and thus better mechanical properties of tube specimens after the ECPT experiment, compared to the outward-discharging scheme. The ECPT can heal microcracks automatically without directly contacting tubular specimens and is not limited by the length of tubular workpieces, exhibiting great potential for crack healing in non-ferrous alloy tubes.

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.

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.

Evaluation of CO2 and CO dopants in hydrogen to reduce hydrogen permeation in the Stirling engine heater head tube alloy CG-27

NASA Technical Reports Server (NTRS)

Misencik, J. A.

1983-01-01

Tubes of CG-27 alloy, filled with hydrogen doped with various amounts of carbon dioxide and carbon monoxide, were heated in a diesel fuel fired Stirling engine simulator materials test rig for 100 hours at 820 C and at a gas pressure of 15 MPa to determine the effectiveness of the dopants in reducing hydrogen permeation through the hot tube wall. This was done for clean as-heat treated tubes and also for tubes that had previously been exposed for 100 hours to hydrogen doped with 1.0 volume percent carbon dioxide to determine if the lower levels of dopant could maintain a low hydrogen permeation through the hot tube wall. Carbon dioxide, as a dopant in hydrogen, was most effective in reducing hydrogen permeation through clean tubes and in maintaining low hydrogen permeation after prior exposure to 1.0 volume percent carbon dioxide. Only the lowest level of carbon dioxide (0.05 volume percent) was not as effective in the clean or prior exposed tubes. Carbon monoxide as a dopant in hydrogen was less effective than carbon dioxide at a given concentration level. Of the four dopant levels studied; 1.0, 0.5, 0.2, and 0.05 volume percent carbon monoxide, only the 1.0 and 0.5 volume percent were effective in reducing and maintaining low hydrogen permeation through the CG-27.

Advanced Ultrasupercritical (AUSC) Tube Membrane Panel Development

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

Pschirer, James; Burgess, Joshua; Schrecengost, Robert

Alstom Power Inc., a wholly owned subsidiary of the General Electric Company (GE), has completed the project “Advanced Ultrasupercritical (AUSC) Tube Membrane Panel Development” under U.S. Department of Energy (DOE) Award Number DE-FE0024076. This project was part of DOE’s Novel Crosscutting Research and Development to Support Advanced Energy Systems program. AUSC Tube Membrane Panel Development was a two and one half year project to develop and verify the manufacturability and serviceability of welded tube membrane panels made from high performance materials suitable for the AUSC steam cycles, defined as high pressure steam turbine inlet conditions of 700-760°C (1292-1400°F) and 24.5-35MPamore » (3500-5000psi). The difficulty of this challenge lies in the fact that the membrane-welded construction imposes demands on the materials that are unlike any that exist in other parts of the boiler. Tube membrane panels have been designed, fabricated, and installed in boilers for over 50 years with relatively favorable experience when fabricated from carbon and Cr-Mo low alloy steels. The AUSC steam cycle requires membrane tube panels fabricated from materials that have not been used in a weldment with metal temperatures in the range of 582-610°C (1080-1130°F). Fabrication materials chosen for the tubing were Grade 92 and HR6W. Grade 92 is a creep strength enhanced ferritic Cr-Mo alloy and HR6W is a high nickel alloy. Once the materials were chosen, GE performed the engineering design of the panels, prepared shop manufacturing drawings, and developed manufacturing and inspection plans. After the materials were purchased, GE manufactured and inspected the tube membrane panels, determined if post fabrication heat treatment of the tube membrane panels was needed, performed pre- and post-weld heat treatment on the Grade 92 panels, conducted final nondestructive inspection of any heat treated tube membrane panels, conducted destructive inspection of the completed

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.

DISSOLUTION OF ZIRCONIUM-CONTAINING FUEL ELEMENTS

DOEpatents

Horn, F.L.

1961-12-12

Uranium is recovered from spent uranium fuel elements containing or clad with zirconium. These fuel elements are placed in an anhydrous solution of hydrogen fluoride and nitrogen dioxide. Within this system uranium forms a soluble complex and zirconium forms an insoluble complex. The uranium can then be separated, treated, and removed from solution as uranium hexafluoride. (AEC)

Processing fissile material mixtures containing zirconium and/or carbon

DOEpatents

Johnson, Michael Ernest; Maloney, Martin David

2013-07-02

A method of processing spent TRIZO-coated nuclear fuel may include adding fluoride to complex zirconium present in a dissolved TRIZO-coated fuel. Complexing the zirconium with fluoride may reduce or eliminate the potential for zirconium to interfere with the extraction of uranium and/or transuranics from fission materials in the spent nuclear fuel.

Nanoindentation study of bulk zirconium hydrides at elevated temperatures

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

Cinbiz, Mahmut Nedim; Balooch, Mehdi; Hu, Xunxiang

Here, the mechanical properties of zirconium hydrides was studied using nano-indentation technique at a temperature range of 25 – 400 °C. Temperature dependency of reduced elastic modulus and hardness of δ- and ε-zirconium hydrides were obtained by conducting nanoindentation experiments on the bulk hydride samples with independently heating capability of indenter and heating stage. The reduced elastic modulus of δ-zirconium hydride (H/Zr ratio =1.61) decreased from ~113 GPa to ~109 GPa while temperature increased from room temperature to 400°C. For ε-zirconium hydrides (H/Zr ratio=1.79), the reduced elastic modulus decreased from 61 GPa to 54 GPa as temperature increased from roommore » temperature to 300 °C. Whereas, hardness of δ-zirconium hydride significantly decreased from 4.1 GPa to 2.41 GPa when temperature increased from room temperature to 400 °C. Similarly, hardness of ε-zirconium hydride decreased from 3.06 GPa to 2.19 GPa with temperature increase from room temperature to 300°C.« less

Nanoindentation study of bulk zirconium hydrides at elevated temperatures

DOE PAGES

Cinbiz, Mahmut Nedim; Balooch, Mehdi; Hu, Xunxiang; ...

2017-08-02

Here, the mechanical properties of zirconium hydrides was studied using nano-indentation technique at a temperature range of 25 – 400 °C. Temperature dependency of reduced elastic modulus and hardness of δ- and ε-zirconium hydrides were obtained by conducting nanoindentation experiments on the bulk hydride samples with independently heating capability of indenter and heating stage. The reduced elastic modulus of δ-zirconium hydride (H/Zr ratio =1.61) decreased from ~113 GPa to ~109 GPa while temperature increased from room temperature to 400°C. For ε-zirconium hydrides (H/Zr ratio=1.79), the reduced elastic modulus decreased from 61 GPa to 54 GPa as temperature increased from roommore » temperature to 300 °C. Whereas, hardness of δ-zirconium hydride significantly decreased from 4.1 GPa to 2.41 GPa when temperature increased from room temperature to 400 °C. Similarly, hardness of ε-zirconium hydride decreased from 3.06 GPa to 2.19 GPa with temperature increase from room temperature to 300°C.« less

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.

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.

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.

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

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

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

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

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.

Effects of the weld thermal cycle on the microstructure of alloy 690

NASA Astrophysics Data System (ADS)

Tuttle, James R.

Alloy 690 has been introduced as a material for use as the heat exchanger tubes in the steam generators (SGs) of pressurised water reactor (PWR) nuclear power plant. Its immediate predecessor, alloy 600, suffered from a number of degradation modes and another alternative, alloy 800, has also had in-service problems. In laboratory tests, alloy 690 in both mill annealed (MA) and special thermally treated (STT) condition has shown a high degree of resistance to degradation in simulated PWR primary side environments and other test media.Limited research has previously been undertaken to investigate the effects of welding on alloy 690, when the material is used in SG applications. It was deemed important to increase knowledge in this area since fabrication of PWR SGs involves gas tungsten arc welding (GTAW) of the heat exchanger tubes to a clad tubeplate. For this research investigation welded samples of alloy 690 have been produced in the laboratory using a range of thermal cycles based around recommended weld parameters for SG fabrication. These samples have been compared with archive welds from PWR SG manufacturers. A number of welds incorporating alloy 600 and a number using alloy 800 tubing material have also been fabricated in the laboratory for comparative purposes. Two experimental melts have been produced to study the effects of Nb substitution for Ti in alloy 690 type materials.Welded and unwelded specimens have been studied, analysed and tested using a variety of methods and techniques. A method of metallographic sample preparation for transmission electron microscope (TEM) thin foil specimens has been developed and documented which ensures foil perforation in a specific region. The effects of Nb substitution for Ti have been discussed. Chemical balances and microstructures in the fusion zone of welds manufactured from alloy 690 tubing incorporating alloy 82 weld consumable have been shown to be non-ideal. Within the heat affected zone (HAZ) of both

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.

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.

Disposition of Chicago Pile 5 (CP-5) Converter Tubes in the 10-160B Cask

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

Pancake, Daniel C.; Rock, Cynthia

This paper will focus on the unique characterization, packaging, and transportation issues associated with the disposition of the two CP-5 Converter Tube assemblies from Argonne National Laboratory. The converter tubes were constructed of combinations of HEU and alloys of zirconium, and were part of the original research facilities attached to the CP-5 reactor during operating evolutions. These assemblies were heavily irradiated during their operational lifetime, and were segregated from the balance of irradiated test specimens when the reactor was deactivated and slated for Decontamination and Demolition (D&D). In addition, the substantial contribution of fissile material to the assemblies’ inventory mademore » the potential disposition pathways extremely challenging. As a result, these items became part of Argonne’s legacy “nuclear footprint”, and were added to the Nuclear Footprint Reduction Project scope for disposition. The Project was responsible for the size reduction and characterization of these items, as well as the ultimate disposition. After negotiating a disposal pathway for these tubes, there were significant transportation issues that required a small team to overcome, in order to successfully ship these items to the Nevada National Security Site (NNSS). The Project team at Argonne, technical support from transportation specialists, licensing support from the 10-160B license owner, the Savanah River National Lab (SRNL) Packaging Certification Team (PCT, and the DOE EM-33 staff contributed to license and safety analysis report amendments that eventually authorized the shipment of the material. The paper will identify the organizations, and the specific actions, required to successfully make three “one of a kind” shipments of irradiated test specimen material. This will include the unique packaging configurations, contents modification for the cask license (via the Amendment process), criticality evaluations, and associated review and

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.

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...

RECOVERY OF URANIUM FROM ZIRCONIUM-URANIUM NUCLEAR FUELS

DOEpatents

Gens, T.A.

1962-07-10

An improvement was made in a process of recovering uranium from a uranium-zirconium composition which was hydrochlorinated with gsseous hydrogen chloride at a temperature of from 350 to 800 deg C resulting in volatilization of the zirconium, as zirconium tetrachloride, and the formation of a uranium containing nitric acid insoluble residue. The improvement consists of reacting the nitric acid insoluble hydrochlorination residue with gaseous carbon tetrachloride at a temperature in the range 550 to 600 deg C, and thereafter recovering the resulting uranium chloride vapors. (AEC)

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.

Grain boundary microstructure, chemistry, and IGSCC in Alloy 600 and Alloy 690

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

Norring, K.; Stiller, K.; Nilsson, J.O.

1992-12-31

The resistance to intergranular stress corrosion cracking of six different Alloy 600 and Alloy 690 steam generator tubes has been investigated. The composition of the materials at grain boundaries has been investigated using analytical transmission electron microscopy and atom probe field ion microscopy techniques. The depletion of chromium at the grain boundaries has been related to the type of grain boundary precipitates. Segregation of carbon and boron to the grain boundaries has been observed and quantified.

The microstructure and precipitation effects in Inconel alloy 690

NASA Astrophysics Data System (ADS)

Smith, Alan J.

Failure of Alloy 600 steam generator tubing in Pressurised Water Reactors (PWRs) has prompted the investigation of alloy 690 as an alternative material. Six commercially produced tubes and ten experimentally produced alloys have been examined with varying amounts of carbon, aluminium and titanium. Alloy compositions have been selected to investigate the individual and combined effects of these elements on the microstructure and corrosion behaviour in the environments of corrosion tests and simulated PWR conditions. Alloys were subjected to simulated mill annealing treatments at varied temperatures. Microstructural characterisation using optical and electron microscopy has demonstrated the effects of composition and thermal treatment in controlling grain size and carbide precipitation together with the interdependence between these structural details. Stress corrosion resistance of selected alloy 690 tubes has been examined with samples in an autoclave at fixed temperatures with environments based on pure water, sodium hydroxide and sodium hydroxide + sodium sulphate solutions. Susceptibility to intergranular attack has been related to aluminium contents of the alloy and to thermal treatments given. Results suggest a decreased resistance to IGA when aluminium is increased. Thermal treatments given to the samples appear not to be very significant to the amounts of IGA. The compositions and heat treatments used in the corrosion study were further examined on a dedicated scanning transmission electron microscope in order to correlate the effects of, chromium depletion, nickel enrichment and impurity segregation at grain boundaries, with corrosion characteristics. These results have shown the effect of varying the special thermal treatment temperature and time on the degree of enrichment / depletion / segregation and the corrosion resistance of the alloy. The mechanism of protection afforded by the special thermal treatment can thus be elucidated.

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.

Cladding burst behavior of Fe-based alloys under LOCA

DOE PAGES

Terrani, Kurt A.; Dryepondt, Sebastien N.; Pint, Bruce A.; ...

2015-12-17

Burst behavior of austenitic and ferritic Fe-based alloy tubes has been examined under a simulated large break loss of coolant accident. Specifically, type 304 stainless steel (304SS) and oxidation resistant FeCrAl tubes were studied alongside Zircaloy-2 and Zircaloy-4 that are considered reference fuel cladding materials. Following the burst test, characterization of the cladding materials was carried out to gain insights regarding the integral burst behavior. Given the widespread availability of a comprehensive set of thermo-mechanical data at elevated temperatures for 304SS, a modeling framework was implemented to simulate the various processes that affect burst behavior in this Fe-based alloy. Themore » most important conclusion is that cladding ballooning due to creep is negligible for Fe-based alloys. Thus, unlike Zr-based alloys, cladding cross-sectional area remains largely unchanged up to the point of burst. Furthermore, for a given rod internal pressure, the temperature onset of burst in Fe-based alloys appears to be simply a function of the alloy's ultimate tensile strength, particularly at high rod internal pressures.« less

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.

Microstructure evolution and mechanical properties degradation of HPNb alloy after a five-year service

NASA Astrophysics Data System (ADS)

Guo, Jingfeng; Cao, Tieshan; Cheng, Congqian; Meng, Xianming; Zhao, Jie

2018-04-01

The microstructure and mechanical properties of ethylene cracking furnace tube (HPNb alloy) are investigated by scanning electronic microscopy (SEM), tensile tests and Charpy impact tests at room temperature, tensile tests and creep tests at high temperature in this paper. The primary carbides of HPNb alloy coarsened and formed a continuous network after a five-year service. Furthermore, a lot of fine secondary carbides precipitated in the dendrite interior. The primary carbides M7C3 and NbC transformed into M23C6 and G phase after service, respectively. The furnace tube after service exhibits higher yield strength, lower tensile strength, worse ductility and toughness than as-cast tube at room temperature. At high temperature, the tensile strength and yield strength of service tube are higher than as-cast tube, but its tensile elongation is lower. The creep strength of HPNb alloy at high temperature decreases after a five-year service. Both microstructure and mechanical properties of ethylene cracking furnace tube have deteriorated after a five-year service.

In vitro assessment of artifacts induced by titanium, titanium-zirconium and zirconium dioxide implants in cone-beam computed tomography.

PubMed

Sancho-Puchades, Manuel; Hämmerle, Christoph H F; Benic, Goran I

2015-10-01

The aim of this study was to test whether or not the intensity of artifacts around implants in cone-beam computed tomography (CBCT) differs between titanium, titanium-zirconium and zirconium dioxide implants. Twenty models of a human mandible, each containing one implant in the single-tooth gap position 45, were cast in dental stone. Five test models were produced for each of the following implant types: titanium 4.1 mm diameter (Ti4.1 ), titanium 3.3 mm diameter (Ti3.3 ), titanium-zirconium 3.3 mm diameter (TiZr3.3 ) and zirconium dioxide 3.5-4.5 mm diameter (ZrO3.5-4.5 ) implants. For control purposes, three models without implants were produced. Each model was scanned using a CBCT device. Gray values (GV) were recorded at eight circumferential positions around the implants at 0.5 mm, 1 mm and 2 mm from the implant surface (GVT est ). GV were assessed in the corresponding volumes of interest (VOI) in the control models without implants (GVC ontrol ). Differences of gray values (ΔGV) between GVT est and GVC ontrol were calculated as percentages. One-way ANOVA and post hoc tests were applied to detect differences between implant types. Mean ΔGV for ZrO3.5-4.5 presented the highest absolute values, generally followed by TiZr3.3 , Ti4.1 and Ti3.3 implants. The differences of ΔGV between ZrO3.5-4.5 and the remaining groups were statistically significant in the majority of the VOI (P ≤ 0.0167). ΔGV for TiZr3.3 , Ti4.1 and Ti3.3 implants did not differ significantly in the most VOI. For all implant types, ΔGV showed positive values buccally, mesio-buccally, lingually and disto-lingually, whereas negative values were detected mesially and distally. Zirconium dioxide implants generate significantly more artifacts as compared to titanium and titanium-zirconium implants. The intensity of artifacts around zirconium dioxide implants exhibited in average the threefold in comparison with titanium implants. © 2014 John Wiley & Sons A/S. Published by John Wiley

Lead-induced stress corrosion cracking of Alloy 600 and 690 in high temperature water

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

Sakai, T.; Senjuh, T.; Aoki, K.

1992-12-31

Lead is one of the potential contributing impurities to the degradation of PWR steam generator tubing. Recent laboratory testing has shown that lead is a corrosive material for Alloy 600 steam generator tubing. However, it is still unknown how lead influences the corrosion of steam generator tubing, including the effect of lead concentration, solution pH, stress level and material characteristics. In this study, two kinds of experiments were performed. One was to investigate the thin film characteristic and selectively dissolved base metal elements of Alloy 600MA in high temperature solutions of different lead concentrations and pH. The other investigated themore » dependency of degradation of Alloy 600MA and Alloy 690TT on lead concentration and stress level in mild acidic environment, at 340{degrees}C for 2500 hrs. It was firstly demonstrated that lead-enhanced selective dissolution of nickel from alloy base metal, as a result of electrochemical reaction between lead and nickel, might cause the initiation and propagation of corrosion. Secondly, we showed that Alloy 690TT, generally very corrosion resistant material, also suffered from Pb-induced corrosion. The difference of the lead-induced stress corrosion morphology of Alloy 600MA and Alloy 690TT was also clarified.« less

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...

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.

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.

Characterization of BOR-60 Irradiated 14YWT-NFA1 Tubes

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

Saleh, Tarik A.; Maloy, Stuart Andrew; Aydogan, Eda

2017-02-15

Tubes of FCRD 14YWT-NFA1 Alloy were placed in the BOR-60 reactor and irradiated under a fast flux neutron environment to two conditions: 7 dpa at 360-370 °C and 6 dpa at 385-430 °C. Small sections of the tube were cut and sent to UC Berkeley for nanohardness testing and focused ion beam (FIB) milling of TEM specimens. FIB specimens were sent back to LANL for final FIB milling and TEM imaging. Hardness data and TEM images are presented in this report. This is the first fast reactor neutron irradiated information on the 14YWT-NFA1 alloy.

Irradiation creep of dispersion strengthened copper alloy

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

Pokrovsky, A.S.; Barabash, V.R.; Fabritsiev, S.A.

1997-04-01

Dispersion strengthened copper alloys are under consideration as reference materials for the ITER plasma facing components. Irradiation creep is one of the parameters which must be assessed because of its importance for the lifetime prediction of these components. In this study the irradiation creep of a dispersion strengthened copper (DS) alloy has been investigated. The alloy selected for evaluation, MAGT-0.2, which contains 0.2 wt.% Al{sub 2}O{sub 3}, is very similar to the GlidCop{trademark} alloy referred to as Al20. Irradiation creep was investigated using HE pressurized tubes. The tubes were machined from rod stock, then stainless steel caps were brazed ontomore » the end of each tube. The creep specimens were pressurized by use of ultra-pure He and the stainless steel caps subsequently sealed by laser welding. These specimens were irradiated in reactor water in the core position of the SM-2 reactors to a fluence level of 4.5-7.1 x 10{sup 21} n/cm{sup 2} (E>0.1 MeV), which corresponds to {approx}3-5 dpa. The irradiation temperature ranged from 60-90{degrees}C, which yielded calculated hoop stresses from 39-117 MPa. A mechanical micrometer system was used to measure the outer diameter of the specimens before and after irradiation, with an accuracy of {+-}0.001 mm. The irradiation creep was calculated based on the change in the diameter. Comparison of pre- and post-irradiation diameter measurements indicates that irradiation induced creep is indeed observed in this alloy at low temperatures, with a creep rate as high as {approx}2 x 10{sup {minus}9}s{sup {minus}1}. These results are compared with available data for irradiation creep for stainless steels, pure copper, and for thermal creep of copper alloys.« less

METHOD OF MAKING DELTA ZIRCONIUM HYDRIDE MONOLITHIC MODERATOR PIECES

DOEpatents

Vetrano, J.B.

1962-01-23

A method is given for preparing large, sound bodies of delta zirconium hydride. The method includes the steps of heating a zirconium body to a temperature of not less than l000 deg C, providing a hydrogen atmosphere for the zirconium body at a pressure not greater than one atmosphere, reducing the temperature slowly to 800 deg C at such a rate that cracks do not form while maintaining the hydrogen pressure substantially constant, and cooling in an atmosphere of hydrogen. (AEC)

Zirconium fluoride glass - Surface crystals formed by reaction with water

NASA Technical Reports Server (NTRS)

Doremus, R. H.; Bansal, N. P.; Bradner, T.; Murphy, D.

1984-01-01

The hydrated surfaces of a zirconium barium fluoride glass, which has potential for application in optical fibers and other optical elements, were observed by scanning electron microscopy. Crystalline zirconium fluoride was identified by analysis of X-ray diffraction patterns of the surface crystals and found to be the main constituent of the surface material. It was also found that hydrated zirconium fluorides form only in highly acidic fluoride solutions. It is possible that the zirconium fluoride crystals form directly on the glass surface as a result of its depletion of other ions. The solubility of zirconium fluoride is suggested to be probably much lower than that of barium fluoride (0.16 g/100 cu cm at 18 C). Dissolution was determined to be the predominant process in the initial stages of the reaction of the glass with water. Penetration of water into the glass has little effect.

Minimalistic Liquid-Assisted Route to Highly Crystalline α-Zirconium Phosphate.

PubMed

Cheng, Yu; Wang, Xiaodong Tony; Jaenicke, Stephan; Chuah, Gaik-Khuan

2017-08-24

Zirconium phosphates have potential applications in areas of ion exchange, catalysis, photochemistry, and biotechnology. However, synthesis methodologies to form crystalline α-zirconium phosphate (Zr(HPO 4 ) 2 ⋅H 2 O) typically involve the use of excess phosphoric acid, addition of HF or oxalic acid and long reflux times or hydrothermal conditions. A minimalistic sustainable route to its synthesis has been developed by using only zirconium oxychloride and concentrated phosphoric acid to form highly crystalline α-zirconium phosphate within hours. The morphology can be changed from platelets to rod-shaped particles by fluoride addition. By varying the temperature and time, α-zirconium phosphate with particle sizes from nanometers to microns can be obtained. Key features of this minimal solvent synthesis are the excellent yields obtained with high atom economy under mild conditions and ease of scalability. © 2017 Wiley-VCH Verlag GmbH & Co. KGaA, Weinheim.

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

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.

International strategic minerals inventory summary report; zirconium

USGS Publications Warehouse

Towner, R.R.

1992-01-01

Zircon, a zirconium silicate, is currently the most important commercial zirconium-bearing mineral. Baddeleyite, a natural form of zirconia, is less important but has some specific end uses. Both zircon and baddeleyite occur in hard-rock and placer deposits, but at present all zircon production is from placer deposits. Most baddeleyite production is from hard-rock deposits, principally as a byproduct of copper and phosphate-rock mining. World zirconium resources in identified, economically exploitable deposits are about 46 times current production rates. Of these resources, some 71 percent are in South Africa, Australia, and the United States. The principal end uses of zirconium minerals are in ceramic applications and as refractories, abrasives, and mold linings in foundries. A minor amount, mainly of zircon, is used for the production of hafnium-free zirconium metal, which is used principally for sheathing fuel elements in nuclear reactors and in the chemical-processing industry, aerospace engineering, and electronics. Australia and South Africa are the largest zircon producers and account for more than 70 percent of world output; the United States and the Soviet Union account for another 20 percent. South Africa accounts for almost all the world's production of baddeleyite, which is about 2 percent of world production of contained zirconia. Australia and South Africa are the largest exporters of zircon. Unless major new deposits are developed in countries that have not traditionally produced zircon, the pattern of world production is unlikely to change by 2020. The proportions, however, of production that come from existing producing countries may change somewhat.

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

Zirconium tetrachloride revisited

DOE PAGES

Borjas Nevarez, Rosendo; Balasekaran, Samundeeswari Mariappan; Kim, Eunja; ...

2018-02-19

We present that zirconium tetrachloride, ZrCl 4, is a strategic material with wide-ranging applications. Until now, only one crystallographic study on ZrCl 4has been reported [Krebs (1970).Z. Anorg. Allg. Chem.378, 263–272] and that was more than 40 years ago. The compound used for the previous determination was prepared from ZrO 2 and Cl 2–CCl 4, and single-crystal X-ray diffraction (SCXRD) studies on ZrCl 4 obtained from Zr metal have not yet been reported. In this context, we prepared ZrCl 4 from the reaction of Zr metal and Cl 2 gas in a sealed tube and investigated its structure at 100,more » 150, 200, 250, and 300 K. At 300 K, the SCXRD analysis indicates that ZrCl 4 crystallizes in the orthorhombic space group Pca2 1 [a= 6.262 (9),b= 7.402 (11),c= 12.039 (17) Å, andV= 558.0 (14) Å 3] and consists of infinite zigzag chains of edge-sharing ZrCl 6 octahedra. This chain motif is similar to that observed previously in ZrCl 4, but the structural parameters and space group differ. Finally, in the temperature range 100–300 K, no phase transformation was identified, while elongation of intra-chain Zr...Zr [3.950 (1) Å at 100 K and 3.968 (5) Å at 300 K] and inter-chain Cl...Cl [3.630 (3) Å at 100 K and 3.687 (9) Å at 300 K] distances occurred.« less

Containerless drop tube solidification and grain refinement of NiAl3

NASA Technical Reports Server (NTRS)

Ethridge, E. C.; Curreri, P. A.; Kelly, M.; Workman, G.; Smith, A. M.; Bond, R.

1984-01-01

The possibility of undercooling Ni-Al alloys below the liquidus in order to produce a single phase peritectic structure by containerless drop tube solidification was studied. Containerless process is a technique for both high purity contamination free studies as well as for investigating the undercooling and rapid solidification of alloys by suppression of heterogeneous nucleation on container walls. In order to achieve large undercoolings one must avoid heterogeneous nucleation of crystallization. It was shown that the Marshall Space Flight Center drop tubes ae unique facilities for containerless solidification experiments and large undercoolings are possible with some alloys. The original goal of undercooling the liquid metal well below the liquidus to the peritectic temperature during containerless free to form primarily NiAl3 was achieved. The microstructures were interesting from another point of view. The microstructure from small diameter samples is greatly refined. Small dendrite arm spacings such as these could greatly facilitate the annealing and solid state transformation of the alloy to nearly 10% NiAl3 by reducing the distance over which diffusion needs to occur. This could minimize annealing time and might make it economically feasible to produce NiAl3 alloy.

Friction Stir Back Extrusion of Aluminium Alloys for Automotive Applications

NASA Astrophysics Data System (ADS)

Xu, Zeren

Since the invention of Friction Stir Welding in 1991 as a solid state joining technique, extensive scientific investigations have been carried out to understand fundamental aspects of material behaviors when processed by this technique, in order to optimize processing conditions as well as mechanical properties of the welds. Based on the basic principles of Friction Stir Welding, several derivatives have also been developed such as Friction Stir Processing, Friction Extrusion and Friction Stir Back Extrusion. Friction Stir Back Extrusion is a novel technique that is proposed recently and designed for fabricating tubes from lightweight alloys. Some preliminary results have been reported regarding microstructure and mechanical properties of Friction Stir Back Extrusion processed AZ31 magnesium alloy, however, systematic study and in-depth investigations are still needed to understand the materials behaviors and underlying mechanisms when subjected to Friction Stir Back Extrusion, especially for age-hardenable Al alloys. In the present study, Friction Stir Back Extrusion processed AA6063-T5 and AA7075-T6 alloys are analyzed with respect to grain structure evolution, micro-texture change, recrystallization mechanisms, precipitation sequence as well as mechanical properties. Optical Microscopy, Electron Backscatter Diffraction, Transmission Electron Microscopy, Vickers Hardness measurements and uniaxial tensile tests are carried out to characterize the microstructural change as well as micro and macro mechanical properties of the processed tubes. Special attention is paid to the micro-texture evolution across the entire tube and dynamic recrystallization mechanisms that are responsible for grain refinement. Significant grain refinement has been observed near the processing zone while the tube wall is characterized by inhomogeneous grain structure across the thickness for both alloys. Dissolution of existing precipitates is noticed under the thermal hysterias imposed by

Thermal treatment, grain boundary composition and intergranular attack resistance of Alloy 690

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

Smith, A.J.; Stratton, R.P.

1992-12-31

Commercial Alloy 690 PWR steam generator tubes and experimentally produced alloys with varying amounts of carbon, aluminium and titanium have been examined. After simulated mill annealing and thermal treatment, the microstructure and corrosion behaviour in corrosion tests have been investigated. Stress corrosion resistance of selected alloy 690 tubes and experimental alloys has been examined with environments based on pure water, sodium hydroxide and sodium hydroxide + sodium sulphate solutions. Effects of aluminium content and the thermal treatments on the susceptibility to intergranular attack have been examined, although they appear not to be very significant to the amounts of IGA. Samplesmore » used in thermal treatments have been further examined with a dedicated scanning transmission electron microscope to show compositional changes at grain boundaries.« less

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

METHOD OF IMPROVING CORROSION RESISTANCE OF ZIRCONIUM

DOEpatents

Shannon, D.W.

1961-03-28

An improved intermediate rinse for zirconium counteracts an anomalous deposit that often results in crevices and outof-the-way places when ordinary water is used to rinse away a strong fluoride etching solution designed to promote passivation of the metal. The intermediate rinse, which is used after the etching solution and before the water, is characterized by a complexing agent for fluoride ions such as aluminum or zirconium nitrates or chlorides.

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.

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...

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:

Direct synthesis of zirconium powder by magnesium reduction

NASA Astrophysics Data System (ADS)

Lee, Dong-Won; Yun, Jung-Yeul; Yoon, Sung-Won; Wang, Jei-Pil

2013-05-01

The direct synthesis of zirconium powder has been conducted through an analysis of the chemical reaction between evaporated ZrCl4 and molten magnesium over a range of reduction temperatures, concentration of hydrochloric acid, and stirring time. The observed results indicated that the purity of zirconium powder increased with increased stirring time, and Mg and MgCl2 were removed by 10 wt% of hydrochloric acid solution. The pure zirconium powder was obtained by stirring again for 5 h using 5 wt% of hydrochloric acid solution. It was noted that the mean particle size increased when the reaction temperature was increased, and the size of the powder at 1,123 K and 1,173 K was found to be 10 μm and 15 μm, respectively. In addition, the purity of the powder was also improved with temperature, and its purity finally reached up to 99.5% at 1,250 K. Overall, pure zirconium powder was obtained after a stirring stage for 5 hours using 5 wt% of hydrochloric acid solution.

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

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.

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.

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.

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.

40 CFR 721.10598 - Lead strontium titanium zirconium oxide.

Code of Federal Regulations, 2013 CFR

2013-07-01

... 40 Protection of Environment 32 2013-07-01 2013-07-01 false Lead strontium titanium zirconium... Specific Chemical Substances § 721.10598 Lead strontium titanium zirconium oxide. (a) Chemical substance and significant new uses subject to reporting. (1) The chemical substance identified as lead strontium...

40 CFR 721.10598 - Lead strontium titanium zirconium oxide.

Code of Federal Regulations, 2014 CFR

2014-07-01

... 40 Protection of Environment 31 2014-07-01 2014-07-01 false Lead strontium titanium zirconium... Specific Chemical Substances § 721.10598 Lead strontium titanium zirconium oxide. (a) Chemical substance and significant new uses subject to reporting. (1) The chemical substance identified as lead strontium...

Fabrication of ultra-fine grained aluminium tubes by RTES technique

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

Jafarzadeh, H., E-mail: h.jafarzadeh@ut.ac.ir; Abrinia, K.

Recently, repetitive tube expansion and shrinking have been exploited as a means for producing ultra-fine grained and nano-crystalline microstructures for magnesium alloy tubes. This method includes two different half-cycles and was based on pressing a tubular part through an angular channel die with two shear zones. Since the aluminium alloys are the most widely used materials in industries, in this study, repetitive tube expansion and shrinking as a new severe plastic deformation technique was applied to commercially pure aluminium for fabricating ultra-fine grained aluminium tubes for the first time and the ability of this process in significant grain refinement ismore » determined even after single cycle. Transmission electron microscopy and X-ray diffraction were used to evaluate the microstructure of the repetitive tube expansion and shrinking processed materials and the examinations showed ultra-fine grains with the average grain size of 320 nm after one cycle of repetitive tube expansion and shrinking. The yield strength, ultimate tensile strength increased notably by the factor of 2.17 and 1.27 respectively, after one cycle of repetitive tube expansion and shrinking, whereas the elongation to failure as well as the uniform elongation decreased. Furthermore, micro-hardness distribution through the part's section proposed the hardness increasing to ~ 55 HV from the initial value of ~ 28 HV after one cycle of repetitive tube expansion and shrinking. - Highlights: • RTES was introduced for fabricating the UFGed AA1050 tubes for the first time. • Nano-grained AA1050 tube was obtained by RTES process. • Grain size of ~ 320 nm was obtained after two half-cycles of RTES process. • Yield and ultimate strength increased by the factor of 2.17 and 1.27 respectively. • The microhardness increased to ~ 55 HV from the initial value of ~ 28 HV.« less

Feasibility of Shape-Memory Ni/Ti Alloy Wire Containing Tube Elevators for Transcrestal Detaching Maxillary Sinus Mucosa: Ex Vivo Study.

PubMed

Li, Yanfeng; Wang, Fuli; Hu, Pin; Fan, Jiadong; Han, Yishi; Liu, Bin; Liu, Tao; Yang, Chunhao; Gu, Xiangmin

2016-01-01

Osteotome sinus floor elevation is a less invasive approach to augment an insufficient alveolar bone at the posterior maxilla for dental implantation. However, this approach has some limitations due to the lack of sinus lift tools available for clinical use and the small transcrestal access to the maxillary sinus floor. We recently invented shape-memory Ni/Ti alloy wire containing tube elevators for transcrestal detaching maxillary sinus mucosa, and developed goat ex vivo models for direct visualizing the effectiveness of detaching sinus mucosa in real time during transcrestal maxillary sinus floor elevation. We evaluated our invented elevators, namely elevator 012 and elevator 014, for their effectiveness for transcrestal detaching maxillary sinus mucosa using the goat ex vivo models. We measured the length of sinus mucosa detached in mesial and distal directions or buccal and palatal directions, and the space volume created by detaching maxillary sinus mucosa in mesial, distal, buccal and palatal directions using the invented elevators. Elevator 012 had a shape-memory Ni/Ti alloy wire with a diameter of 0.012 inch, while elevator 014 had its shape-memory Ni/Ti alloy wire with a diameter of 0.014 inch. Elevator 012 could detach the goat maxillary sinus mucosa in the mesial or distal direction for 12.1±4.3 mm, while in the buccal or palatal direction for 12.5±6.7 mm. The elevator 014 could detach the goat maxillary sinus mucosa for 23.0±4.9 mm in the mesial or distal direction, and for 19.0±8.1 mm in the buccal or palatal direction. An average space volume of 1.7936±0.2079 ml was created after detaching the goat maxillay sinus mucosa in both mesial/distal direction and buccal/palatal direction using elevator 012; while the average space volume created using elevator 014 was 1.8764±0.2366 ml. Both two newly invented tube elevators could effectively detach the maxillary sinus mucosa on the goat ex vivo sinus models. Moreover, elevator 014 has advantages over

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.

High-Temperature Crystal-Growth Cartridge Tubes Made by VPS

NASA Technical Reports Server (NTRS)

Holmes, Richard; O'Dell, Scott; McKechnie, Timothy; Power, Christopher

2008-01-01

Cartridge tubes for use in a crystal growth furnace at temperatures as high as 1,600 deg. C have been fabricated by vacuum plasma spraying (VPS). These cartridges consist mainly of an alloy of 60 weight percent molybdenum with 40 weight percent rhenium, made from molybdenum powder coated with rhenium. This alloy was selected because of its high melting temperature (approximately equal.2,550 C) and because of its excellent ductility at room temperature. These cartridges are intended to supplant tungsten/nickel-alloy cartridges, which cannot be used at temperatures above approximately equal 1,300 C.

Geologic structure of Gofitsky deposit of titanium and zirconium and perspectives of the reserve base of titanium and zirconium in Russia

NASA Astrophysics Data System (ADS)

Kukhmazov, Iskander

2016-04-01

With the fall of the Soviet Union, all the mining deposits of titanium and zirconium appeared outside of Russian Federation. Therefore the studying of deposits of titanium and zirconium in Russia is very important nowadays. There is a paradoxical situation in the country: in spite of possible existence of national mineral resource base of Ti-Zr material, which can cover needs of the country, Russia is the one of the largest buyers of imported Ti-Zr material in the world. Many deposits are not mined, and those which are in the process of mining have poor reserves. Demand for this raw material is very great not only for Russia, but also for the world in general. Today there is a scarcity of zircon around the world and it will only increase through time. Therefore prices of products of titanium and zirconium also increase. Consequently Russian deposits of titanium and zirconium with higher content than foreign may become competitive. Russia is forced to buy raw materials (zirconium and titanium production) from former Soviet Union countries at prices higher than the world's and thus incur huge losses, including customs charges. Russia should create its own mineral resource base of Ti-Zr. Studied titanium-zirconium deposits of Stavropol region may become the basis for the south part of Russia. At first, Beshpagirsky deposit should be pointed out. It has large reserves of ore sands with high content of Ti-Zr. A combination of favorable geographical position of the area with developed industrial infrastructure makes it very beneficial as an object for high priority development. Gofitsky deposit should be pointed out as well. Its sands have a wide areal distribution and a high content of titanium and zirconium. Chokrak, Karagan-Konksk and Sarmatian sediments of the Miocene of Gofitsky deposit are productive for titanium and zirconium placers within Stavropol region of Russia. Gofitsky deposit was evaluated from financial and economic point of view and the following data

Thermal expansion method for lining tantalum alloy tubing with tungsten

NASA Technical Reports Server (NTRS)

Watson, G. K.; Whittenberger, J. D.; Mattson, W. F.

1973-01-01

A differential-thermal expansion method was developed to line T-111 (tantalum - 8 percent tungsten - 2 percent hafnium) tubing with a tungsten diffusion barrier as part of a fuel element fabrication study for a space power nuclear reactor concept. This method uses a steel mandrel, which has a larger thermal expansion than T-111, to force the tungsten against the inside of the T-111 tube. Variables investigated include lining temperature, initial assembly gas size, and tube length. Linear integrity increased with increasing lining temperature and decreasing gap size. The method should have more general applicability where cylinders must be lined with a thin layer of a second material.

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

Biaxial Thermal Creep of Alloy 617 and Alloy 230 for VHTR Applications

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

Mo, Kun; Lv, Wei; Tung, Hsiao-Ming

2016-05-18

In this study, we employed pressurized creep tubes to investigate the biaxial thermal creep behavior of Inconel 617 (alloy 617) and Haynes 230 (alloy 230). Both alloys are considered to he the primary candidate structural materials for very high-temperature reactors (VITITRs) due to their exceptional high-temperature mechanical properties. The current creep experiments were conducted at 900 degrees C for the effective stress range of 15-35 MPa. For both alloys, complete creep strain development with primary, secondary, and tertiary regimes was observed in all the studied conditions. Tertiary creep was found to he dominant over the entire creep lives of bothmore » alloys. With increasing applied creep stress, the fraction of the secondary creep regime decreases. The nucleation, diffusion, and coarsening of creep voids and carbides on grain boundaries were found to be the main reasons for the limited secondary regime and were also found to be the major causes of creep fracture. The creep curves computed using the adjusted creep equation of the form epsilon= cosh 1(1 rt) + P-sigma ntm agree well with the experimental results for both alloys at die temperatures of 850-950 degrees C.« less

Electroslag Strip Cladding of Steam Generators With Alloy 690

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

Consonni, M.; Maggioni, F.; Brioschi, F.

2006-07-01

The present paper details the results of electroslag cladding and tube-to-tubesheet welding qualification tests conducted by Ansaldo-Camozzi ESC with Alloy 690 (Alloy 52 filler metal) on steel for nuclear power stations' steam generators shell, tubesheet and head; the possibility of submerged arc cladding on first layer was also considered. Test results, in terms of chemical analysis, mechanical properties and microstructure are reproducible and confidently applicable to production cladding and show that electroslag process can be used for Alloy 52 cladding with exceptionally stable and regular operation and high productivity. The application of submerged arc cladding process to the first layermore » leads to a higher base metal dilution, which should be avoided. Moreover, though the heat affected zone is deeper with electroslag cladding, in both cases no coarsened grain zone is found due to recrystallization effect of second cladding layer. Finally, the application of electroslag process to cladding of Alloy 52 with modified chemical composition, was proved to be highly beneficial as it strongly reduces hot cracking sensitivity, which is typical of submerged arc cladded Alloy 52, both during tube-to-tubesheet welding and first re-welding. (authors)« 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.

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.

METHOD OF PREPARING SINTERED ZIRCONIUM METAL FROM ITS HYDRIDES

DOEpatents

Angier, R.P.

1958-02-11

The invention relates to the preparation of metal shapes from zirconium hydride by powder metallurgical techniques. The zirconium hydride powder which is to be used for this purpose can be prepared by rendering massive pieces of crystal bar zirconium friable by heat treatment in purified hydrogen. This any then be ground into powder and powder can be handled in the air without danger of it igniting. It may then be compacted in the normal manner by being piaced in a die. The compact is sintered under vacuum conditions preferably at a temperature ranging from 1200 to 1300 deg C and for periods of one to three hours.

Texture and hydride orientation relationship of Zircaloy-4 fuel clad tube during its fabrication for pressurized heavy water reactors

NASA Astrophysics Data System (ADS)

Vaibhaw, Kumar; Rao, S. V. R.; Jha, S. K.; Saibaba, N.; Jayaraj, R. N.

2008-12-01

Zircaloy-4 material is used for cladding tube in pressurized heavy water reactors (PHWRs) of 220 MWe and 540 MWe capacity in India. These tubes are fabricated by using various combinations of thermo-mechanical processes to achieve desired mechanical and corrosion properties. Cladding tube develops crystallographic texture during its fabrication, which has significant influence on its in-reactor performance. Due to radiolytic decomposition of water Zircaloy-4 picks-up hydrogen. This hydrogen in excess of its maximum solubility in reactor operating condition (˜300 °C), precipitates as zirconium hydrides causing embrittlement of cladding tube. Hydride orientation in the radial direction of the tube limits the service life and lowers the fuel burn-up in reactor. The orientation of the hydride primarily depends on texture developed during fabrication. A correlation between hydride orientation ( F n) with the texture in the tube during its fabrication has been developed using a second order polynomial. The present work is aimed at quantification and correlation of texture evolved in Zircaloy-4 cladding tube using Kearn's f-parameter during its fabrication process.

Braze material for joining ceramic to metal and ceramic to ceramic surfaces and joined ceramic to metal and ceramic to ceramic article

DOEpatents

Hunt, T.K.; Novak, R.F.

1991-05-07

An improved active metal braze filler material is provided in which the coefficient of thermal expansion of the braze filler is more closely matched with that of the ceramic and metal, or two ceramics, to provide ceramic to metal, or ceramic to ceramic, sealed joints and articles which can withstand both high temperatures and repeated thermal cycling without failing. The braze filler material comprises a mixture of a material, preferably in the form of a powder, selected from the group consisting of molybdenum, tungsten, silicon carbide and mixtures thereof, and an active metal filler material selected from the group consisting of alloys or mixtures of nickel and titanium, alloys or mixtures of nickel and zirconium, alloys or mixtures of nickel, titanium, and copper, alloys or mixtures of nickel, titanium, and zirconium, alloys or mixtures of niobium and nickel, alloys or mixtures of niobium and zirconium, alloys or mixtures of niobium and titanium, alloys or mixtures of niobium, titanium, and nickel, alloys or mixtures of niobium, zirconium, and nickel, and alloys or mixtures of niobium, titanium, zirconium, and nickel. The powder component is selected such that its coefficient of thermal expansion will effect the overall coefficient of thermal expansion of the braze material so that it more closely matches the coefficients of thermal expansion of the ceramic and metal parts to be joined. 3 figures.

Braze material for joining ceramic to metal and ceramic to ceramic surfaces and joined ceramic to metal and ceramic to ceramic article

DOEpatents

Hunt, Thomas K.; Novak, Robert F.

1991-01-01

An improved active metal braze filler material is provided in which the coefficient of thermal expansion of the braze filler is more closely matched with that of the ceramic and metal, or two ceramics, to provide ceramic to metal, or ceramic to ceramic, sealed joints and articles which can withstand both high temperatures and repeated thermal cycling without failing. The braze filler material comprises a mixture of a material, preferably in the form of a powder, selected from the group consisting of molybdenum, tungsten, silicon carbide and mixtures thereof, and an active metal filler material selected from the group consisting of alloys or mixtures of nickel and titanium, alloys or mixtures of nickel and zirconium, alloys or mixtures of nickel, titanium, and copper, alloys or mixtures of nickel, titanium, and zirconium, alloys or mixtures of niobium and nickel, alloys or mixtures of niobium and zirconium, alloys or mixtures of niobium and titanium, alloys or mixtures of niobium, titanium, and nickel, alloys or mixtures of niobium, zirconium, and nickel, and alloys or mixtures of niobium, titanium, zirconium, and nickel. The powder component is selected such that its coefficient of thermal expansion will effect the overall coefficient of thermal expansion of the braze material so that it more closely matches the coefficients of thermal expansion of the ceramic and metal parts to be joined.

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

Mechanical resistance of zirconium implant abutments: A review of the literature

PubMed Central

Vaquero-Aguilar, Cristina; Torres-Lagares, Daniel; Jiménez-Melendo, Manuel; Gutiérrez-Pérez, José L.

2012-01-01

The increase of aesthetic demands, together with the successful outcome of current implants, has renewed interest in the search for new materials with enough mechanical properties and better aesthetic qualities than the materials customarily used in implanto-prosthetic rehabilitation. Among these materials, zirconium has been used in different types of implants, including prosthetic abutments. The aim of the present review is to analyse current scientific evidence supporting the use of this material for the above mentioned purposes. We carried out the review of the literature published in the last ten years (2000 through 2010) of in vitro trials of dynamic and static loading of zirconium abutments found in the databases of Medline and Cochrane using the key words zirconium abutment, fracture resistance, fracture strength, cyclic loading. Although we have found a wide variability of values among the different studies, abutments show favourable clinical behaviour for the rehabilitation of single implants in the anterior area. Such variability may be explained by the difficulty to simulate daily mastication under in vitro conditions. The clinical evidence, as found in our study, does not recommend the use of implanto-prosthetic zirconium abutments in the molar area. Key words: Zirconium abutment, zirconium implant abutment, zirconia abutment, fracture resistance, fracture strength, cyclic loading. PMID:22143702

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...

Characterization of Friction Stir Welded Tubes by Means of Tube Bulge Test

NASA Astrophysics Data System (ADS)

D'Urso, G.; Longo, M.; Giardini, C.

2011-05-01

Mechanical properties of friction stir welded joints are generally evaluated by means of conventional tensile test. This testing method might provide insufficient information because maximum strain obtained in tensile test before necking is small; moreover, the application of tensile test is limited when the joint path is not linear or even when the welds are executed on curved surfaces. Therefore, in some cases, it would be preferable to obtain the joints properties from other testing methods. Tube bulge test can be a valid solution for testing circumferential or longitudinal welds executed on tubular workpieces. The present work investigates the mechanical properties and the formability of friction stir welded tubes by means of tube bulge tests. The experimental campaign was performed on tubular specimens having a thickness of 3 mm and an external diameter of 40 mm, obtained starting from two semi-tubes longitudinally friction stir welded. The first step, regarding the fabrication of tubes, was performed combining a conventional forming process and friction stir welding. Sheets in Al-Mg-Si-Cu alloy AA6060 T6 were adopted for this purpose. Plates having a dimension of 225×60 mm were bent (with a bending axis parallel to the main dimension) in order to obtain semi-tubes. A particular care was devoted to the fabrication of forming devices (punch and die) in order to minimize the springback effects. Semi-tubes were then friction stir welded by means of a CNC machine tool. Some preliminary tests were carried out by varying the welding parameters, namely feed rate and rotational speed. A very simple tool having flat shoulder and cylindrical pin was used. The second step of the research was based on testing the welded tubes by means of tube bulge test. A specific equipment having axial actuators with a conical shape was adopted for this study. Some analyses were carried out on the tubes bulged up to a certain pressure level. In particular, the burst pressure and the

The Study of Titanium and Zirconium Ions in Water by MPT-LTQ Mass Spectrometry in Negative Mode

PubMed Central

Yang, Junqing; Zheng, Mei; Liu, Qiuju; Zhu, Meiling; Yang, Chushan; Zhang, Yan; Zhu, Zhiqiang

2017-01-01

Microwave plasma torches (MPTs) can be used as simple and low power-consumption ambient ion sources. When MPT-mass spectrometry (MPT-MS) is applied in the detection of some metal elements, the metallic ions exhibit some novel features which are significantly different with those obtained by the traditional inductively coupled plasma (ICP)-mass spectrometry (ICP-MS) and may be helpful for metal element analysis. As the representative elements of group IVA, titanium and zirconium are both of importance and value in modern industry, and they have impacts on human health. Here, we first provide a study on the complex anions of titanium and zirconium in water by using the MPT as ion source and a linear ion trap mass spectrometer (LTQ-MS). These complex anions were produced in the plasma flame by an aqueous solution flowing through the central tube of the MPT, and were introduced into the inlet of the mass spectrometry working in negative ion mode to get the feature mass spectrometric signals. Moreover, the feature fragment patterns of these ions in multi-step collision- induced dissociation processes have been explained. Under the optimized conditions, the limit of detection (LOD) using the MS2 (the second tandem mass spectrometry) procedure was estimated to be at the level of 10 μg/L for titanium and 20 μg/L for zirconium with linear dynamics ranges that cover at least two orders of magnitude, i.e., between 0–500 μg/L and 20–200 μg/L, respectively. These experimental data demonstrated that the MPT-MS is a promising and useful tool in field analysis of titanium and zirconium ions in water, and can be applied in many fields, such as environmental control, hydrogeology, and water quality inspection. In addition, MPT-MS could also be used as a supplement of ICP-MS for the rapid and on-site analysis of metal ions. PMID:28954404

The Study of Titanium and Zirconium Ions in Water by MPT-LTQ Mass Spectrometry in Negative Mode.

PubMed

Yang, Junqing; Zheng, Mei; Liu, Qiuju; Yang, Meiling Zhu Chushan; Zhang, Yan; Zhu, Zhiqiang

2017-09-26

Microwave plasma torches (MPTs) can be used as simple and low power-consumption ambient ion sources. When MPT-mass spectrometry (MPT-MS) is applied in the detection of some metal elements, the metallic ions exhibit some novel features which are significantly different with those obtained by the traditional inductively coupled plasma (ICP)-mass spectrometry (ICP-MS) and may be helpful for metal element analysis. As the representative elements of group IVA, titanium and zirconium are both of importance and value in modern industry, and they have impacts on human health. Here, we first provide a study on the complex anions of titanium and zirconium in water by using the MPT as ion source and a linear ion trap mass spectrometer (LTQ-MS). These complex anions were produced in the plasma flame by an aqueous solution flowing through the central tube of the MPT, and were introduced into the inlet of the mass spectrometry working in negative ion mode to get the feature mass spectrometric signals. Moreover, the feature fragment patterns of these ions in multi-step collision- induced dissociation processes have been explained. Under the optimized conditions, the limit of detection (LOD) using the MS² (the second tandem mass spectrometry) procedure was estimated to be at the level of 10μg/L for titanium and 20 μg/L for zirconium with linear dynamics ranges that cover at least two orders of magnitude, i.e., between 0-500 μg/L and 20-200 μg/L, respectively. These experimental data demonstrated that the MPT-MS is a promising and useful tool in field analysis of titanium and zirconium ions in water, and can be applied in many fields, such as environmental control, hydrogeology, and water quality inspection. In addition, MPT-MS could also be used as a supplement of ICP-MS for the rapid and on-site analysis of metal ions.

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

77 FR 41967 - Certain Circular Welded Carbon Steel Pipes and Tubes From India, Thailand, and Turkey; Certain...

Federal Register 2010, 2011, 2012, 2013, 2014

2012-07-17

... Pipes and Tubes From India, Thailand, and Turkey; Certain Circular Welded Non-Alloy Steel Pipe From... on (1) certain circular welded carbon steel pipes and tubes from India, Thailand, and Turkey; (2... circular welded carbon steel pipes and tubes from Turkey would likely lead to continuation or recurrence of...

Thermochemistry of amorphous and crystalline zirconium and hafnium silicates.

NASA Astrophysics Data System (ADS)

Ushakov, S.; Brown, C. E.; Navrotsky, Alexandra; Boatner, L. A.; Demkov, A. A.; Wang, C.; Nguyen, B.-Y.

2003-03-01

Calorimetric investigation of amorphous and crystalline zirconium and hafnium silicates was performed as part of a research program on thermochemistry of alternative gate dielectrics. Amorphous hafnium and zirconium silicates with varying SiO2 content were synthesized by a sol-gel process. Crystalline zirconium and hafnium silicates (zircon and hafnon) were synthesized by solid state reaction at 1450 °C from amorphous gels and grown as single crystals from flux. High temperature oxide melt solution calorimetry in lead borate (2PbO.B2O3) solvent at 800 oC was used to measure drop solution enthalpies for amorphous and crystalline zirconium and hafnium silicates and corresponding oxides. Applying appropriate thermochemical cycles, formation enthalpy of crystalline ZrSiO4 (zircon) from binary oxides (baddeleite and quartz) at 298 K was calculated as -23 +/-2 kJ/mol and enthalpy difference between amorphous and crystalline zirconium silicate (vitrification enthalpy) was found to be 61 +/-3 kJ/mol. Crystallization onset temperatures of amorphous zirconium and hafnium silicates, as measured by differential scanning calorimetry (DSC), increased with silica content. The resulting crystalline phases, as characterized by X-ray diffraction (XRD), were tetragonal HfO2 and ZrO2. Critical crystallite size for tetragonal to monoclinic transformation of HfO2 in the gel was estimated as 6 +/-2 nm from XRD data Crystallization enthalpies per mole of hafnia and zirconia in gels decrease slightly together with crystallite size with increasing silica content, for example from -22 to -15 +/-1 kJ per mol of HfO2 crystallized at 740 and 1006 °C from silicates with 10 and 70 mol Applications of thermal analyses and solution calorimetry techniques together with first-principles density functional calculations to estimate interface and surface energies are discussed.

[The clinical application of zirconium-dioxide-ceramics. Case report].

PubMed

Somfai, Dóra; Zsigmond, Ágnes; Károlyházy, Katalin; Kispély, Barbara; Hermann, Péter

2015-12-01

Due to its outstanding physical, mechanical and esthetic properties, zirconium-dioxide is one of the most popular non-metal denture, capable of surpassing PFM in most cases. The recent advances of CAD/CAM technology makes it a good alternitve. Here we show the usefulness of zirconium-dioxide in everyday dental practice through three case reports.

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.

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.

Overcoming the crystallization and designability issues in the ultrastable zirconium phosphonate framework system

DOE PAGES

Zheng, Tao; Yang, Zaixing; Gui, Daxiang; ...

2017-05-30

Metal-organic frameworks (MOFs) based on zirconium phosphonates exhibit superior chemical stability suitable for applications under harsh conditions. These compounds mostly exist as poorly crystallized precipitates, and precise structural information has therefore remained elusive. Furthermore, a zero-dimensional zirconium phosphonate cluster acting as secondary building unit has been lacking, leading to poor designability in this system. We overcome these challenges and obtain single crystals of three zirconium phosphonates that are suitable for structural analysis. Furthermore, these compounds are built by previously unknown isolated zirconium phosphonate clusters and exhibit combined high porosity and ultrastability even in fuming acids. SZ-2 possesses the largest voidmore » volume recorded in zirconium phosphonates and SZ-3 represents the most porous crystalline zirconium phosphonate and the only porous MOF material reported to survive in aqua regia. SZ-2 and SZ-3 can effectively remove uranyl ions from aqueous solutions over a wide pH range, and we have elucidated the removal mechanism.« less

Comparison of Zirconium Phosphonate-Modified Surfaces for Immobilizing Phosphopeptides and Phosphate-Tagged Proteins.

PubMed

Forato, Florian; Liu, Hao; Benoit, Roland; Fayon, Franck; Charlier, Cathy; Fateh, Amina; Defontaine, Alain; Tellier, Charles; Talham, Daniel R; Queffélec, Clémence; Bujoli, Bruno

2016-06-07

Different routes for preparing zirconium phosphonate-modified surfaces for immobilizing biomolecular probes are compared. Two chemical-modification approaches were explored to form self-assembled monolayers on commercially available primary amine-functionalized slides, and the resulting surfaces were compared to well-characterized zirconium phosphonate monolayer-modified supports prepared using Langmuir-Blodgett methods. When using POCl3 as the amine phosphorylating agent followed by treatment with zirconyl chloride, the result was not a zirconium-phosphonate monolayer, as commonly assumed in the literature, but rather the process gives adsorbed zirconium oxide/hydroxide species and to a lower extent adsorbed zirconium phosphate and/or phosphonate. Reactions giving rise to these products were modeled in homogeneous-phase studies. Nevertheless, each of the three modified surfaces effectively immobilized phosphopeptides and phosphopeptide tags fused to an affinity protein. Unexpectedly, the zirconium oxide/hydroxide modified surface, formed by treating the amine-coated slides with POCl3/Zr(4+), afforded better immobilization of the peptides and proteins and efficient capture of their targets.

Overcoming the crystallization and designability issues in the ultrastable zirconium phosphonate framework system

PubMed Central

Zheng, Tao; Yang, Zaixing; Gui, Daxiang; Liu, Zhiyong; Wang, Xiangxiang; Dai, Xing; Liu, Shengtang; Zhang, Linjuan; Gao, Yang; Chen, Lanhua; Sheng, Daopeng; Wang, Yanlong; Diwu, Juan; Wang, Jianqiang; Zhou, Ruhong; Chai, Zhifang; Albrecht-Schmitt, Thomas E.; Wang, Shuao

2017-01-01

Metal-organic frameworks (MOFs) based on zirconium phosphonates exhibit superior chemical stability suitable for applications under harsh conditions. These compounds mostly exist as poorly crystallized precipitates, and precise structural information has therefore remained elusive. Furthermore, a zero-dimensional zirconium phosphonate cluster acting as secondary building unit has been lacking, leading to poor designability in this system. Herein, we overcome these challenges and obtain single crystals of three zirconium phosphonates that are suitable for structural analysis. These compounds are built by previously unknown isolated zirconium phosphonate clusters and exhibit combined high porosity and ultrastability even in fuming acids. SZ-2 possesses the largest void volume recorded in zirconium phosphonates and SZ-3 represents the most porous crystalline zirconium phosphonate and the only porous MOF material reported to survive in aqua regia. SZ-2 and SZ-3 can effectively remove uranyl ions from aqueous solutions over a wide pH range, and we have elucidated the removal mechanism. PMID:28555656

Overcoming the crystallization and designability issues in the ultrastable zirconium phosphonate framework system

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

Zheng, Tao; Yang, Zaixing; Gui, Daxiang

Metal-organic frameworks (MOFs) based on zirconium phosphonates exhibit superior chemical stability suitable for applications under harsh conditions. These compounds mostly exist as poorly crystallized precipitates, and precise structural information has therefore remained elusive. Furthermore, a zero-dimensional zirconium phosphonate cluster acting as secondary building unit has been lacking, leading to poor designability in this system. We overcome these challenges and obtain single crystals of three zirconium phosphonates that are suitable for structural analysis. Furthermore, these compounds are built by previously unknown isolated zirconium phosphonate clusters and exhibit combined high porosity and ultrastability even in fuming acids. SZ-2 possesses the largest voidmore » volume recorded in zirconium phosphonates and SZ-3 represents the most porous crystalline zirconium phosphonate and the only porous MOF material reported to survive in aqua regia. SZ-2 and SZ-3 can effectively remove uranyl ions from aqueous solutions over a wide pH range, and we have elucidated the removal mechanism.« less

Zirconium diselenite microstructures, formation and mechanism

NASA Astrophysics Data System (ADS)

Naik, Chandan C.; Salker, A. V.

2018-04-01

In this work, a series of microstructures of zirconium diselenite (Zr(SeO3)2) has been prepared via a simple precipitation method at room temperature without adding any organic surfactants. Phase purity of the sample has been checked by X-ray Diffraction. From the SEM, FESEM, and TEM images spheroid nanoparticles to the starfish-like structure of zirconium diselenite are detected. The morphological evolution processes were investigated carefully following time-dependent experiments and a growth mechanism has been proposed. Two different crystal growth processes, the oriented attachment process accompanying the Ostwald ripening process were held responsible for the formation of a structure resembling starfish having four arms.

Characterization of mechanical properties of hydroxyapatite-silicon-multi walled carbon nano tubes composite coatings synthesized by EPD on NiTi alloys for biomedical application.

PubMed

Khalili, Vida; Khalil-Allafi, Jafar; Sengstock, Christina; Motemani, Yahya; Paulsen, Alexander; Frenzel, Jan; Eggeler, Gunther; Köller, Manfred

2016-06-01

Release of Ni(1+) ions from NiTi alloy into tissue environment, biological response on the surface of NiTi and the allergic reaction of atopic people towards Ni are challengeable issues for biomedical application. In this study, composite coatings of hydroxyapatite-silicon multi walled carbon nano-tubes with 20wt% Silicon and 1wt% multi walled carbon nano-tubes of HA were deposited on a NiTi substrate using electrophoretic methods. The SEM images of coated samples exhibit a continuous and compact morphology for hydroxyapatite-silicon and hydroxyapatite-silicon-multi walled carbon nano-tubes coatings. Nano-indentation analysis on different locations of coatings represents the highest elastic modulus (45.8GPa) for HA-Si-MWCNTs which is between the elastic modulus of NiTi substrate (66.5GPa) and bone tissue (≈30GPa). This results in decrease of stress gradient on coating-substrate-bone interfaces during performance. The results of nano-scratch analysis show the highest critical distance of delamination (2.5mm) and normal load before failure (837mN) as well as highest critical contact pressure for hydroxyapatite-silicon-multi walled carbon nano-tubes coating. The cell culture results show that human mesenchymal stem cells are able to adhere and proliferate on the pure hydroxyapatite and composite coatings. The presence of both silicon and multi walled carbon nano-tubes (CS3) in the hydroxyapatite coating induce more adherence of viable human mesenchymal stem cells in contrast to the HA coated samples with only silicon (CS2). These results make hydroxyapatite-silicon-multi walled carbon nano-tubes a promising composite coating for future bone implant application. Copyright © 2016 Elsevier Ltd. All rights reserved.

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.

40 CFR 721.10602 - Lead niobium titanium zirconium oxide.

Code of Federal Regulations, 2013 CFR

2013-07-01

... as specified in § 721.90 (a)(4), (b)(4), and (c)(4) (Where N=8, and 8 is an aggregate of releases for the following substances: Lead strontium titanium zirconium oxide (PMN P-11-270; CAS No. 61461-40-3... strontium titanium tungsten oxide (PMN P-11-272; CAS No. 1262279-30-0); Lanthanum lead titanium zirconium...

40 CFR 721.10602 - Lead niobium titanium zirconium oxide.

Code of Federal Regulations, 2014 CFR

2014-07-01

... as specified in § 721.90 (a)(4), (b)(4), and (c)(4) (Where N=8, and 8 is an aggregate of releases for the following substances: Lead strontium titanium zirconium oxide (PMN P-11-270; CAS No. 61461-40-3... strontium titanium tungsten oxide (PMN P-11-272; CAS No. 1262279-30-0); Lanthanum lead titanium zirconium...

Surface characterization of anodized zirconium for biomedical applications

NASA Astrophysics Data System (ADS)

Sanchez, A. Gomez; Schreiner, W.; Duffó, G.; Ceré, S.

2011-05-01

Mechanical properties and corrosion resistance of zirconium make this material suitable for biomedical implants. Its good in vivo performance is mainly due to the presence of a protective oxide layer that minimizes corrosion rate, diminishes the amount of metallic ions released to the biological media and facilitates the osseointegration process. Since the implant surface is the region in contact with living tissues, the characteristics of the surface film are of great interest. Surface modification is a route to enhance both biocompatibility and corrosion resistance of permanent implant materials. Anodizing is presented as an interesting process to modify metal surfaces with good reproducibility and independence of the geometry. In this work the surface of zirconium before and after anodizing in 1 mol/L phosphoric acid solution at a fixed potential between 3 and 30 V, was characterized by means of several surface techniques. It was found that during anodization the surface oxide grows with an inhomogeneous coverage on zirconium surface, modifying the topography. The incorporation of P from the electrolyte to the surface oxide during the anodizing process changes the surface chemistry. After 30 days of immersion in Simulated Body Fluid (SBF) solution, Ca-P rich compounds were present on anodized zirconium.

Specific Adaptation of Gas Atomization Processing for Al-Based Alloy Powder for Additive Manufacturing

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

Anderson, Iver; Siemon, John

The initial three atomization attempts resulted in “freeze-outs” within the pour tubes in the pilot-scale system and yielded no powder. Re-evaluation of the alloy liquidus temperatures and melting characteristics, in collaboration with Alcoa, showed further superheat to be necessary to allow the liquid metal to flow through the pour tube to the atomization nozzle. A subsequent smaller run on the experimental atomization system verified these parameters and was successful, as were all successive runs on the larger pilot scale system. One alloy composition froze-out part way through the atomization on both pilot scale runs. SEM images showed needle formation andmore » phase segregations within the microstructure. Analysis of the pour tube freeze-out microstructures showed that large needles formed within the pour tube during the atomization experiment, which eventually blocked the melt stream. Alcoa verified the needle formation in this alloy using theoretical modeling of phase solidification. Sufficient powder of this composition was still generated to allow powder characterization and additive manufacturing trials at Alcoa.« less

SEPARATION OF URANIUM FROM ZIRCONIUM AND NIOBIUM BY SOLVENT EXTRACTION

DOEpatents

Voiland, E.E.

1958-05-01

A process for separation of the uranium from zirconium and/or niobium values contained in 3 to 7M aqueous nitric acid solutions is described. This is accomplished by adding phosphoric acid anions to the nitric acid solution containing the uranium, zirconium, and/or niobium in an amount sufficient to make the solution 0.05 to 0.2M in phosphate ion and contacting the solution with an organic water-immiscible solvent such as MEK, whereby the uranyl values are taken up by the extract phase while the zirconium and niobium preferentially remain in the aqueous raffinate.

In-situ stabilization of radioactive zirconium swarf

DOEpatents

Hess, Clay C.

1999-01-01

The method for treating ignitable cutting swarf in accordance with the present invention involves collecting cutting swarf in a casting mold underwater and injecting a binder mixture comprising vinyl ester styrene into the vessel to fill void volume; and form a mixture comprising swarf and vinyl ester styrene; and curing the mixture. The method is especially useful for stabilizing the ignitable characteristics of radioactive zirconium cutting swarf, and can be used to solidify zirconium swarf, or other ignitable finely divided material, underwater. The process could also be performed out of water with other particulate wastes.

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.

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.

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.

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.

Upper critical fields and superconducting transition temperatures of some zirconium-base amorphous transition-metal alloys

NASA Astrophysics Data System (ADS)

Karkut, M. G.; Hake, R. R.

1983-08-01

Superconducting upper critical fields Hc2(T), transition temperatures Tc and normal-state electrical resistivities ρn have been measured in the amorphous transition-metal alloy series Zr1-xCox, Zr1-xNix, (Zr1-xTix)0.78Ni0.22, and (Zr1-xNbx)0.78Ni0.22. Structural integrity of these melt-spun alloys is indicated by x-ray, density, bend-ductility, normal-state electrical resistivity, superconducting transition width, and mixed-state flux-pinning measurements. The specimens display Tc=2.1-3.8 K, ρn=159-190 μΩ cm, and |(dHc2dT)Tc|=28-36 kG/K. These imply electron mean free paths l~2-6 Å, zero-temperature Ginzburg-Landau coherence distances ξG0~50-70 Å, penetration depths λG0~(7-10)×103 Å, and extremely high dirtiness parameters ξ0l~300-1300. All alloys display Hc2(T) curves with negative curvature and (with two exceptions) fair agreement with the standard dirty-limit theory of Werthamer, Helfand, Hohenberg, and Maki (WHHM) for physically reasonable values of spin-orbit-coupling induced, electron-spin-flip scattering time τso. This is in contrast to the anomalously elevated Hc2(T) behavior which is nearly linear in T that is observed by some, and the unphysically low-τso fits to WHHM theory obtained by others, for various amorphous alloys. Current ideas that such anomalies may be due to alloy inhomogeneity are supported by present results on two specimens for which relatively low-τso fits of Hc2(T) to WHHM theory are coupled with superconductive evidence for inhomogeneity: relatively broad transitions at Tc and Hc2 current-density-dependent transitions at Hc2 and (in one specimen) a J-dependent, high-H (>Hc2), resistive "beak effect." In the Zr1-xCox and Zr1-xNix series, Tc decreases linearly with x (and with unfilled-shell average electron-to-atom ratio < ea > in the range 5.05<=< ea ><=6.40 in fair agreement with previous results for these systems and contrary to the Tc vs < ea > behavior of both amorphous and crystalline transition-metal alloys formed

Permanent and separable aerospace tubing/ fitting evaluation program, volume 1

NASA Technical Reports Server (NTRS)

Ursell, C. R.; Franz, C. R.

1975-01-01

The tube fitting evaluation program was conducted to develop a standard overall test program including methods, procedures, and equipment as well as recommended test sequences for qualifying fitting/tubing assemblies. The program consisted of testing the MS flareless (separable) fitting and utilizing the results as baseline data from which all other fittings will be evaluated. Five separable designs and five permanent designs were tested in three sizes with two types of tubing materials. The basic design requirements were 4,000 psi operating pressure within the temperature range of minus 65 F to plus 450 F while also being compatible with 21-6-9 and titanium 3Al-2.5V tubing alloys.

Role of microstructure in caustic stress corrosion cracking of Alloy 690

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

Mertz, D.A.; Duda, P.T.; Pica, P.N.

1995-12-31

Alloy 690 has been selected for nuclear heat transport system tubing application in recent commercial reactor plants due to its resistance to multiple types of corrosion attack. Typical corn final heat treatments for this material are a mill-anneal (MA, approximately 1,070 C) to completely dissolve the carbides and develop the final grain structure plus a thermal treatment (TT, approximately 700 C) to precipitate carbides at the grain boundaries. Tubing with grain boundary carbides and no or few intragranular carbides has been found resistant to intergranular stress corrosion cracking (IGSCC) in caustic environments. In this work, first, Alloy 690 plate wasmore » subjected to a variety of MA and MA-TT heat treatments to create microstructures of carbide-decorated grain boundaries and undecorated boundaries. Caustic IGSCC test results were consistent with tubing data. Second, experiments were conducted to understand the mechanism by which caustic-corrosion resistance is imparted to Alloy 690 by grain boundary carbides. Tubing with a fully-developed MA-TT carbide microstructure was strained and heat-treated to create a mixed microstructure of new grain boundaries with no carbide precipitate decoration, intermixed with intragranular carbide strings from prior grain boundaries. Caustic SCC performance of this material was identical to that of material with the MA-TT carbide-decorated grain boundaries. This work suggests that the fundamental cause of good IGSCC resistance of MA-TT Alloy 690 in caustic does not derive solely from grain boundary carbides. It is suggested that matrix strength, as measured by yield stress, could be a controlling factor.« less

Electron-emission characteristics of tungsten alloys: Mee 492. [No data; plasma anode tube

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

Moon, H.M.

A plasma-anode tube was constructed to investigate the electron-emission characteristics for rhenium, 1% thorium tungsten, and 2% thorium tungsten. The tube consists of cathode, anode, cesium reservoir, and three probes. Inside of tube is a rough vacuum by using a Varian liquid-nitrogen-cooled cryovalve and further vacuum by using a Varian absorption pumpt to 4 X IO/sup -5/ pa. The tube was sealed off from the vacuum pump after the cesium ampoule was broken. The entire plasma-anode tube except the cesiunm reservoir was placed in a Blue M Electric Company furnace whose door had been modified to permit viewing of themore » tube.« less

Hopping conduction in zirconium oxynitrides thin film deposited by reactive magnetron sputtering

NASA Astrophysics Data System (ADS)

Guo, Jie; Zhan, Guanghui; Liu, Jingquan; Yang, Bin; Xu, Bin; Feng, Jie; Chen, Xiang; Yang, Chunsheng

2015-10-01

Zirconium oxynitrides thin film thermometers were demonstrated to be useful temperature sensors. However, the basic conduction mechanism of zirconium oxynitrides films has been a long-standing issue, which hinders the prediction and optimization of their ultimate performance. In this letter, zirconium oxynitrides films were grown on sapphire substrates by magnetron sputtering and their electric transport mechanism has been systemically investigated. It was found that in high temperatures region (>150 K) the electrical conductivity was dominated by thermal activation for all samples. In the low temperatures range, while Mott variable hopping conduction (VRH) was dominated the transport for films with relatively low resistance, a crossover from Mott VRH conduction to Efros-Shklovskii (ES) VRH was observed for films with relatively high resistance. This low temperature crossover from Mott to ES VRH indicates the presence of a Coulomb gap (~7 meV). These results demonstrate the competing and tunable conduction mechanism in zirconium oxynitrides thin films, which would be helpful for optimizing the performance of zirconium oxynitrides thermometer.

Isomerization of Cyclooctadiene to Cyclooctyne with a Zinc/Zirconium Heterobimetallic Complex

PubMed Central

Butler, Michael J.; White, Andrew J. P.

2016-01-01

Abstract Reaction of a zinc/zirconium heterobimetallic complex with 1,5‐cyclooctadiene (1,5‐COD) results in slow isomerization to 1,3‐cyclooctadiene (1,3‐COD), along with the formation of a new complex that includes a cyclooctyne ligand bridging two metal centers. While analogous magnesium/zirconium and aluminum/zirconium heterobimetallic complexes are competent for the catalytic isomerization of 1,5‐COD to 1,3‐COD, only in the case of the zinc species is the cyclooctyne adduct observed. PMID:27071992

Lithium aluminate/zirconium material useful in the production of tritium

DOEpatents

Cawley, W.E.; Trapp, T.J.

A composition is described useful in the production of tritium in a nuclear reactor. Lithium aluminate particles are dispersed in a matrix of zirconium. Tritium produced by the reactor of neutrons with the lithium are absorbed by the zirconium, thereby decreasing gas pressure within capsules carrying the material.

Lithium aluminate/zirconium material useful in the production of tritium

DOEpatents

Cawley, W.E.; Trapp, T.J.

1984-10-09

A composition is described useful in the production of tritium in a nuclear reactor. Lithium aluminate particles are dispersed in a matrix of zirconium. Tritium produced by the reactor of neutrons with the lithium are absorbed by the zirconium, thereby decreasing gas pressure within capsules carrying the material.

Lithium aluminate/zirconium material useful in the production of tritium

DOEpatents

Cawley, William E.; Trapp, Turner J.

1984-10-09

A composition is described useful in the production of tritium in a nuclear eactor. Lithium aluminate particles are dispersed in a matrix of zirconium. Tritium produced by the reactor of neutrons with the lithium are absorbed by the zirconium, thereby decreasing gas pressure within capsules carrying the material.

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.

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.

Laser Brazing of High Temperature Braze Alloy

NASA Technical Reports Server (NTRS)

Gao, Y. P.; Seaman, R. F.; McQuillan, T. J.; Martiens, R. F.

2000-01-01

The Space Shuttle Main Engine (SSME) consists of 1080 conical tubes, which are furnace brazed themselves, manifolds, and surrounding structural jacket making almost four miles of braze joints. Subsequent furnace braze cycles are performed due to localized braze voids between the coolant tubes. SSME nozzle experiences extremely high heat flux (180 mW/sq m) during hot fire. Braze voids between coolant tubes may result in hot combustion gas escape causing jacket bulges. The nozzle can be disqualified for flight or result in mission failure if the braze voids exceed the limits. Localized braze processes were considered to eliminate braze voids, however, damage to the parent materials often prohibited use of such process. Being the only manned flight reusable rocket engine, it has stringent requirement on the braze process. Poor braze quality or damage to the parent materials limits the nozzle service life. The objective of this study was to develop a laser brazing process to provide quality, localized braze joints without adverse affect on the parent materials. Gold (Au-Cu-Ni-Pd-Mn) based high temperature braze alloys were used in both powder and wire form. Thin section iron base superalloy A286 tube was used as substrate materials. Different Laser Systems including CO2 (10.6 micrometers, 1kW), ND:YAG (1.06 micrometers, 4kW). and direct diode laser (808nm. 150W) were investigated for brazing process. The laser process variables including wavelength. laser power, travel speed and angle of inclination were optimized according to bead geometry and braze alloy wetting at minimum heat input level, The properties of laser brazing were compared to that of furnace brazing. Microhardness profiles were used for braze joint property comparison between laser and furnace brazing. The cooling rate of laser brazing was compared to furnace brazing based on secondary dendritic arm spacing, Both optical and Scanning Electron Microscope (SEM) were used to evaluate the microstructures of

Thermal storage/discharge performances of Cu-Si alloy for solar thermochemical process

NASA Astrophysics Data System (ADS)

Gokon, Nobuyuki; Yamaguchi, Tomoya; Cho, Hyun-seok; Bellan, Selvan; Hatamachi, Tsuyoshi; Kodama, Tatsuya

2017-06-01

The present authors (Niigata University, Japan) have developed a tubular reactor system using novel "double-walled" reactor/receiver tubes with carbonate molten-salt thermal storage as a phase change material (PCM) for solar reforming of natural gas and with Al-Si alloy thermal storage as a PCM for solar air receiver to produce high-temperature air. For both of the cases, the high heat capacity and large latent heat (heat of solidification) of the PCM phase circumvents the rapid temperature change of the reactor/receiver tubes at high temperatures under variable and uncontinuous characteristics of solar radiation. In this study, we examined cyclic properties of thermal storage/discharge for Cu-Si alloy in air stream in order to evaluate a potentiality of Cu-Si alloy as a PCM thermal storage material. Temperature-increasing performances of Cu-Si alloy are measured during thermal storage (or heat-charge) mode and during cooling (or heat-discharge) mode. A oxidation state of the Cu-Si alloy after the cyclic reaction was evaluated by using electron probe micro analyzer (EPMA).

Equations of state for crystalline zirconium iodide: The role of dispersion

NASA Astrophysics Data System (ADS)

Rossi, Matthew L.; Taylor, Christopher D.

2013-02-01

We present the first-principle equations of state of several zirconium iodides, ZrI2, ZrI3, and ZrI4, computed using density functional theory methods that apply various methods for introducing the dispersion correction. Iodides formed due to reaction of molecular or atomic iodine with zirconium and zircaloys are of particular interest due to their application to the cladding material used in the fabrication of nuclear fuel rods. Stress corrosion cracking (SCC), associated with fission product chemistry with the clad material, is a major concern in the life cycle of nuclear fuels, as many of the observed rod failures have occurred due to pellet-cladding chemical interactions (PCCI) [A. Atrens, G. Dannhäuser, G. Bäro, Stress-corrosion-cracking of zircaloy-4 cladding tubes, Journal of Nuclear Materials 126 (1984) 91-102; P. Rudling, R. Adamson, B. Cox, F. Garzarolli, A. Strasser, High burn-up fuel issues, Nuclear Engineering and Technology 40 (2008) 1-8]. A proper understanding of the physical properties of the corrosion products is, therefore, required for the development of a comprehensive SCC model. In this particular work, we emphasize that, while existing modeling techniques include methods to compute crystal structures and associated properties, it is important to capture intermolecular forces not traditionally included, such as van der Waals (dispersion) correction. Furthermore, crystal structures with stoichiometries favoring a high I:Zr ratio are found to be particularly sensitive, such that traditional density functional theory approaches that do not incorporate dispersion incorrectly predict significantly larger volumes of the lattice. This latter point is related to the diffuse nature of the iodide electron cloud.

Isomerization of Cyclooctadiene to Cyclooctyne with a Zinc/Zirconium Heterobimetallic Complex.

PubMed

Butler, Michael J; White, Andrew J P; Crimmin, Mark R

2016-06-06

Reaction of a zinc/zirconium heterobimetallic complex with 1,5-cyclooctadiene (1,5-COD) results in slow isomerization to 1,3-cyclooctadiene (1,3-COD), along with the formation of a new complex that includes a cyclooctyne ligand bridging two metal centers. While analogous magnesium/zirconium and aluminum/zirconium heterobimetallic complexes are competent for the catalytic isomerization of 1,5-COD to 1,3-COD, only in the case of the zinc species is the cyclooctyne adduct observed. © 2016 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

PHYSICAL PROPERTIES OF ZIRCONIUM NITRIDE IN THE HOMOGENEITY REGION (in Ukrainian)

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

Samsonov, G.V.; Verkhoglyadova, T.S.

1962-01-01

The x-ray method was used to determine the homogeneity region of zirconium nitride as 40 to 50 at.% (9.5 to 13.3% by weight) of nitrogen. It is also shown that part of the ionic bond in the zirconium nitride lattice increases with a decrease in the nitrogen content in this region, this increase being higher than in the homogeneity region of titunium nitride due to the smaller degree of unfilling of the electron d-shell of the zirconium atom in comparison with that of the titanium atom. (auth)

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.

Surface conditions of Nitinol wires, tubing, and as-cast alloys. The effect of chemical etching, aging in boiling water, and heat treatment.

PubMed

Shabalovskaya, S A; Anderegg, J; Laab, F; Thiel, P A; Rondelli, G

2003-04-15

The surface conditions of Nitinol wires and tubing were evaluated with the use of X-ray photoelectron spectroscopy, high-resolution Auger spectroscopy, electron backscattering, and scanning-electron microscopy. Samples were studied in the as-received state as well as after chemical etching, aging in boiling water, and heat treatment, and compared to a mechanically polished 600-grit-finish Nitinol surface treated similarly. General regularities in surface behavior induced by the examined surface treatments are similar for wires, tubing, and studied as-cast alloy, though certain differences in surface Ni concentration were observed. Nitinol wires and tubing from various suppliers demonstrated great variability in Ni surface concentration (0.5-15 at.%) and Ti/Ni ratio (0.4-35). The wires in the as-received state, with the exception of those with a black oxide originating in the processing procedure, revealed nickel and titanium on the surface in both elemental and oxidized states, indicating a nonpassive surface. Shape-setting heat treatment at 500 degrees C for 15 min resulted in tremendous increase in the surface Ni concentration and complete Ni oxidation. Preliminary chemical etching and boiling in water successfully prevented surface enrichment in Ni, initially resulting from heat treatment. A stoichiometric uniformly amorphous TiO(2) oxide generated during chemical etching and aging in boiling water was reconstructed at 700 degrees C, revealing rutile structure. Copyright 2003 Wiley Periodicals, Inc.

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.

In-situ stabilization of radioactive zirconium swarf

DOEpatents

Hess, C.C.

1999-08-31

The method for treating ignitable cutting swarf in accordance with the present invention involves collecting cutting swarf in a casting mold underwater and injecting a binder mixture comprising vinyl ester styrene into the vessel to fill void volume; and form a mixture comprising swarf and vinyl ester styrene; and curing the mixture. The method is especially useful for stabilizing the ignitable characteristics of radioactive zirconium cutting swarf, and can be used to solidify zirconium swarf, or other ignitable finely divided material, underwater. The process could also be performed out of water with other particulate wastes. 6 figs.

Electrochemical Interpretation of a Stress Corrosion Cracking of Thermally Treated Ni base Alloys in a Lead Contaminated Water

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

Hwang, Seong Sik; Lim, Yun Soo; Kim, Hong Pyo

2007-08-20

Since the PbSCC(Lead stress corrosion cracking) of alloy 600 tubing materials was reported by Copson and Dean in 1965, the effect of lead on a corrosion film and cracking morphology have been continually debated. An electrochemical interaction of lead with the alloying elements of SG tubings was studied and the corrosion products were analyzed. It was found that lead enhanced the anodic dissolution of alloy 600 and alloy 690 in the electrochemical test. The lead preferentially dissolved the Cr from the corrosion film of alloy 600 and alloy 690 in alkaline water. The lead ion seemed to penetrate into themore » TG crack tip and react with the corrosion film. A selective Cr depletion was observed to weaken the stability of the passive film on the alloys. Whereas passivity of Ni became stable in lead containing solution, Cr and Fe passivity became unstable.« less

PROCESS OF PREPARING ZIRCONIUM OXYCHLORIDE

DOEpatents

Wilhelm, H.A.; Andrews, M.L.

1960-06-28

A process is given for preparing zirconyl chloride by mixing solid zirconyl chloride octahydrate and solid zirconium tetrachloride at room temperature whereby both chlorides are converted to zirconyl chloride trinydrate and hydrogen chloride is formed and volatilized by the reaction heat.

Recovery of aluminium, nickel-copper alloys and salts from spent fluorescent lamps.

PubMed

Rabah, Mahmoud A

2004-01-01

This study explores a combined pyro-hydrometallurgical method to recover pure aluminium, nickel-copper alloy(s), and some valuable salts from spent fluorescent lamps (SFLs). It also examines the safe recycling of clean glass tubes for the fluorescent lamp industry. Spent lamps were decapped under water containing 35% acetone to achieve safe capture of mercury vapour. Cleaned glass tubes, if broken, were cut using a rotating diamond disc to a standard shorter length. Aluminium and copper-nickel alloys in the separated metallic parts were recovered using suitable flux to decrease metal losses going to slag. Operation variables affecting the quality of the products and the extent of recovery with the suggested method were investigated. Results revealed that total loss in the glass tube recycling operation was 2% of the SFLs. Pure aluminium meeting standard specification DIN 1712 was recovered by melting at 800 degrees C under sodium chloride/carbon flux for 20 min. Standard nickel-copper alloys with less than 0.1% tin were prepared by melting at 1250 degrees C using a sodium borate/carbon flux. De-tinning of the molten nickel-copper alloy was carried out using oxygen gas. Tin in the slag as oxide was recovered by reduction using carbon or hydrogen gas at 650-700 degrees C. Different valuable chloride salts were also obtained in good quality. Further research is recommended on the thermodynamics of nickel-copper recovery, yttrium and europium recovery, and process economics.

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.

Creep-rupture behavior of seven iron-base alloys after long term aging at 760 deg in low pressure hydrogen

NASA Technical Reports Server (NTRS)

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

1980-01-01

Seven candidate iron-base alloys for heater tube application in the Stirling automotive engine were aged for 3500 hours at 760 C in argon and hydrogen. Aging degraded the tensile and creep-rupture properties. The presence of hydrogen during aging caused additional degradiation of the rupture strength in fine grain alloys. Based on current design criteria for the Mod 1 Stirling engine, N-155 and 19-9DL are considered the only alloys in this study with strengths adequate for heater tube service at 760 C.

Corrosion inhibition of steam generator tubesheet by Alloy 690 cladding in secondary side environments

NASA Astrophysics Data System (ADS)

Hur, Do Haeng; Choi, Myung Sik; Lee, Deok Hyun; Han, Jung Ho; Shim, Hee Sang

2013-11-01

Denting is a phenomenon that a steam generator tube is distorted by a volume expansion of corrosion products of the tube support and tubesheet materials adjacent to the tube. Although denting has been mitigated by a modification of the design and material of the tube support structures, it has been an inevitable concern in the crevice region of the top of tubesheet. This paper provides a new technology to prevent denting by cladding the secondary surface of the tubesheet with a corrosion resistant material. In this study, Alloy 690 material was cladded onto the surface of an SA508 tubesheet to a thickness of about 9 mm. The corrosion rates of the original SA508 tubesheet and the Alloy 690 clad material were measured in acidic and alkaline simulated environments. Using Alloy 690 cladding, the corrosion rate of the tubesheet within a magnetite sludge pile decreased by a factor of 680 in 0.1 M NiCl2 solution at 300 °C, and by a factor of 58 in 2 M NaOH solution at 315 °C. This means that denting can drastically be prevented by cladding the secondary tubesheet surface with corrosion resistant materials.

Intercalation chemistry of zirconium 4-sulfophenylphosphonate

NASA Astrophysics Data System (ADS)

Svoboda, Jan; Zima, Vítězslav; Melánová, Klára; Beneš, Ludvík; Trchová, Miroslava

2013-12-01

Zirconium 4-sulfophenylphosphonate is a layered material which can be employed as a host for the intercalation reactions with basic molecules. A wide range of organic compounds were chosen to represent intercalation ability of zirconium 4-sulfophenylphosphonate. These were a series of alkylamines from methylamine to dodecylamine, 1,4-phenylenediamine, p-toluidine, 1,8-diaminonaphthalene, 1-aminopyrene, imidazole, pyridine, 4,4‧-bipyridine, poly(ethylene imine), and a series of amino acids from glycine to 6-aminocaproic acid. The prepared compounds were characterized by powder X-ray diffraction, thermogravimetry analysis and IR spectroscopy and probable arrangement of the guest molecules in the interlayer space of the host is proposed based on the interlayer distance of the prepared intercalates and amount of the intercalated guest molecules.

Microstructure and mechanical properties of FeCrAl alloys under heavy ion irradiations

NASA Astrophysics Data System (ADS)

Aydogan, E.; Weaver, J. S.; Maloy, S. A.; El-Atwani, O.; Wang, Y. Q.; Mara, N. A.

2018-05-01

FeCrAl ferritic alloys are excellent cladding candidates for accident tolerant fuel systems due to their high resistance to oxidation as a result of formation of a protective Al2O3 scale at high temperatures in steam. In this study, we report the irradiation response of the 10Cr and 13Cr FeCrAl cladding tubes under Fe2+ ion irradiation up to ∼16 dpa at 300 °C. Dislocation loop size, density and characteristics were determined using both two-beam bright field transmission electron microscopy and on-zone scanning transmission electron microscopy techniques. 10Cr (C06M2) tube has a lower dislocation density, larger grain size and a slightly weaker texture compared to the 13Cr (C36M3) tube before irradiation. After irradiation to 0.7 dpa and 16 dpa, the fraction of <100> type sessile dislocations decreases with increasing Cr amount in the alloys. It has been found that there is neither void formation nor α‧ precipitation as a result of ion irradiations in either alloy. Therefore, dislocation loops were determined to be the only irradiation induced defects contributing to the hardening. Nanoindentation testing before the irradiation revealed that the average nanohardness of the C36M3 tube is higher than that of the C06M2 tube. The average nanohardness of irradiated tube samples saturated at 1.6-2.0 GPa hardening for both tubes between ∼3.4 dpa and ∼16 dpa. The hardening calculated based on transmission electron microscopy was found to be consistent with nanohardness measurements.

Microstructure and mechanical properties of FeCrAl alloys under heavy ion irradiations

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

Aydogan, E.; Weaver, J. S.; Maloy, S. A.

FeCrAl ferritic alloys are excellent cladding candidates for accident tolerant fuel systems due to their high resistance to oxidation as a result of formation of a protective Al 2O 3 scale at high temperatures in steam. In this study, we report the irradiation response of the 10Cr and 13Cr FeCrAl cladding tubes under Fe 2+ ion irradiation up to ~16 dpa at 300 °C. Dislocation loop size, density and characteristics were determined using both two beam bright field transmission electron microscopy and on-zone scanning transmission electron microscopy techniques. 10Cr (C06M2) tube has a lower dislocation density, larger grain size andmore » a slightly weaker texture compared to the 13Cr (C36M3) tube before irradiation. After irradiation to 0.7 dpa and 16 dpa, the fraction of <100> type sessile dislocations decreases with increasing Cr amount in the alloys. It has been found that there is neither void formation nor α' precipitation as a result of ion irradiations in either alloy. Therefore, dislocation loops were determined to be the only irradiation induced defects contributing to the hardening. Nanoindentation testing before the irradiation revealed that the average nanohardness of the C36M3 tube is higher than that of the C06M2 tube. The average nanohardness of irradiated tube samples saturated at 1.6-2.0 GPa hardening for both tubes between ~3.4 dpa and ~16 dpa. The hardening calculated based on transmission electron microscopy was found to be consistent with nanohardness measurements.« less

Microstructure and mechanical properties of FeCrAl alloys under heavy ion irradiations

DOE PAGES

Aydogan, E.; Weaver, J. S.; Maloy, S. A.; ...

2018-03-02

FeCrAl ferritic alloys are excellent cladding candidates for accident tolerant fuel systems due to their high resistance to oxidation as a result of formation of a protective Al 2O 3 scale at high temperatures in steam. In this study, we report the irradiation response of the 10Cr and 13Cr FeCrAl cladding tubes under Fe 2+ ion irradiation up to ~16 dpa at 300 °C. Dislocation loop size, density and characteristics were determined using both two beam bright field transmission electron microscopy and on-zone scanning transmission electron microscopy techniques. 10Cr (C06M2) tube has a lower dislocation density, larger grain size andmore » a slightly weaker texture compared to the 13Cr (C36M3) tube before irradiation. After irradiation to 0.7 dpa and 16 dpa, the fraction of <100> type sessile dislocations decreases with increasing Cr amount in the alloys. It has been found that there is neither void formation nor α' precipitation as a result of ion irradiations in either alloy. Therefore, dislocation loops were determined to be the only irradiation induced defects contributing to the hardening. Nanoindentation testing before the irradiation revealed that the average nanohardness of the C36M3 tube is higher than that of the C06M2 tube. The average nanohardness of irradiated tube samples saturated at 1.6-2.0 GPa hardening for both tubes between ~3.4 dpa and ~16 dpa. The hardening calculated based on transmission electron microscopy was found to be consistent with nanohardness measurements.« less

Thermodynamic Analysis and Growth of Zirconium Carbide by Chemical Vapor Deposition

NASA Astrophysics Data System (ADS)

Wei, Sun; Hua, Hao Zheng; Xiang, Xiong

Equilibrium calculations were used to optimize conditions for the chemical vapor deposition of zirconium carbide from zirconium halide + CxHy+H2+Ar system. The results show the CVD-ZrC phase diagram is divided into ZrC+C, ZrC and ZrC+Zr zones by C, Zr generating lines. For the same mole of ZrCl4 reactant, it needs higher concentration of CH4 to generate single ZrC phase than that of C3H6. Using these calculations as a guide, single-phase cubic zirconium carbide coatings were deposited onto graphite substrate.

Microstructure evolution and tensile properties of Zr-2.5wt%Nb pressure tubes processed from billets with different microstructures

NASA Astrophysics Data System (ADS)

Kapoor, K.; Saratchandran, N.; Muralidharan, K.

1999-02-01

Starting with identical ingots, billets having different microstructures were obtained by three different processing methods for fabrication of Zr-2.5wt%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 structure 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.

Extractive separation of uranium and zirconium sulfates by amines

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

Schroetterova, D.; Nekovar, P.; Mrnka, M.

1992-04-01

This paper describes an amine extraction process for zirconium and uranium separation. The behaviour of an extraction system containing uranium (VI) sulfate, zirconium (IV) sulfate, 0.2 and 0.5 M sulfuric acid (as the original aqueous phase), tertiary amine tri-n-lauryl- amine or primary amine Primene JMT in benzene (as the original organic phase) is discussed on the basis of equilibrium data. The measured dependences show that the degree of extraction of zirconium at the sulfuric acid concentration of 0.5 M and above is only slightly affected by a presence of uranium in solution. From this surprising behaviour it follows that zirconiummore » may be employed for the displacement of uranium from the organic phase. This effect is more pronounced with the primary amine Primene JMT than with TLA. 29 refs., 4 figs., 1 tab.« less

21 CFR 700.16 - Use of aerosol cosmetic products containing zirconium.

Code of Federal Regulations, 2013 CFR

2013-04-01

... 21 Food and Drugs 7 2013-04-01 2013-04-01 false Use of aerosol cosmetic products containing... SERVICES (CONTINUED) COSMETICS GENERAL Requirements for Specific Cosmetic Products § 700.16 Use of aerosol cosmetic products containing zirconium. (a) Zirconium-containing complexes have been used as an ingredient...

21 CFR 700.16 - Use of aerosol cosmetic products containing zirconium.

Code of Federal Regulations, 2012 CFR

2012-04-01

... 21 Food and Drugs 7 2012-04-01 2012-04-01 false Use of aerosol cosmetic products containing... SERVICES (CONTINUED) COSMETICS GENERAL Requirements for Specific Cosmetic Products § 700.16 Use of aerosol cosmetic products containing zirconium. (a) Zirconium-containing complexes have been used as an ingredient...

21 CFR 700.16 - Use of aerosol cosmetic products containing zirconium.

Code of Federal Regulations, 2014 CFR

2014-04-01

... 21 Food and Drugs 7 2014-04-01 2014-04-01 false Use of aerosol cosmetic products containing... SERVICES (CONTINUED) COSMETICS GENERAL Requirements for Specific Cosmetic Products § 700.16 Use of aerosol cosmetic products containing zirconium. (a) Zirconium-containing complexes have been used as an ingredient...

From Zirconium Nanograins to Zirconia Nanoneedles

PubMed Central

Zalnezhad, E.; Hamouda, A. M. S.; Jaworski, J.; Do Kim, Young

2016-01-01

Combinations of three simple techniques were utilized to gradually form zirconia nanoneedles from zirconium nanograins. First, a physical vapor deposition magnetron sputtering technique was used to deposit pure zirconium nanograins on top of a substrate. Second, an anodic oxidation was applied to fabricate zirconia nanotubular arrays. Finally, heat treatment was used at different annealing temperatures in order to change the structure and morphology from nanotubes to nanowires and subsequently to nanoneedles in the presence of argon gas. The size of the pure zirconium nanograins was estimated to be approximately 200–300 nm. ZrO2 nanotubular arrays with diameters of 70–120 nm were obtained. Both tetragonal and monoclinic ZrO2 were observed after annealing at 450 °C and 650 °C. Only a few tetragonal peaks appeared at 850 °C, while monoclinic ZrO2 was obtained at 900 °C and 950 °C. In assessing the biocompatibility of the ZrO2 surface, the human cell line MDA-MB-231 was found to attach and proliferate well on surfaces annealed at 850 °C and 450 °C; however, the amorphous ZrO2 surface, which was not heat treated, did not permit extensive cell growth, presumably due to remaining fluoride. PMID:27623486

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.

Translucency and Strength of High-Translucency Monolithic Zirconium-Oxide Materials

DTIC Science & Technology

2016-05-12

APPROV~, Col Drew W. Fallis Dean, Air Force Postgraduate Dental School r UNIFORMED SERVICES UNIVERSITY OF THE HEALTH SCIENCES AIR FORCE...POSTGRADUATE DENTAL SCHOOL 2450 Pepperrell Street Lackland AFB Texas, 78236-5345 http://www.usuhs.mil "The author hereby certifies that the use of any...Translucency Monolithic Zirconium-Oxide Materials Abstract Dental materials manufacturers have developed more translucent monolithic zirconium oxide

Microstructural characterization and mechanical properties of Excel alloy pressure tube material

NASA Astrophysics Data System (ADS)

Sattari, Mohammad

Microstructural characterization and mechanical properties of Excel (Zr-3.5%Sn-0.8%Mo-0.8%Nb), a dual phase alphaZr -hcp and betaZr-bcc pressure tube material, is discussed in the current study which is presented in manuscript format. Chapter 3 discusses phase transformation temperatures using different techniques such as quantitative metallography, differential scanning calorimetry (DSC), and electrical resistivity. It was found that the alphaZr → alphaZr+beta Zr and alphaZr+betaZr → betaZr transformation temperatures are in the range of 600-690°C and 960-970°C respectively. Also it was observed that upon quenching from temperatures below ˜860°C the martensitic transformation of betaZr to alpha'--hcp is halted and instead the microstructure transforms into retained Zr with o hexagonal precipitates inside betaZr grains. Chapter 4 deals with aging response of Excel alloy. Precipitation hardening was observed in samples water-quenched from high in the alphaZr+beta Zr or betaZr regions followed by aging. The optimum aging conditions were found to be 450°C for 1 hour. Transmission electron microscopy (TEM) showed dispersion of fine precipitates (˜10nm) inside the martensitic phase. Energy dispersive X-ray spectroscopy (EDS) showed the chemical composition of precipitates to be Zr-30wt%Mo-25wt%Nb-2wt%Fe. Electron crystallography using whole pattern symmetry of the convergent beam electron diffraction (CBED) patterns together with selected area diffraction (SAD) polycrystalline ring patterns, suggests the -6m2 point group for the precipitates belonging to hexagonal crystal structure, with a= 2.936 A and c=4.481 A, i.e. c/a =1.526. Crystallographic texture and high temperature tensile properties as well as creep-rupture properties of different microstructures are discussed in Chapter 5. Texture analysis showed that solution treatment high in the alpha Zr+betaZr or betaZr regions followed by water quenching or air cooling results in a more random texture compared

Determination of fluoride in water - A modified zirconium-alizarin method

USGS Publications Warehouse

Lamar, W.L.

1945-01-01

A convenient, rapid colorimetric procedure using the zirconium-alizarin indicator acidified with sulfuric acid for the determination of fluoride in water is described. Since this acid indicator is stable indefinitely, it is more useful than other zirconium-alizarin reagents previously reported. The use of sulfuric acid alone in acidifying the zirconium-alizarin reagent makes possible the maximum suppression of the interference of sulfate. Control of the pH of the samples eliminates errors due to the alkalinity of the samples. The fluoride content of waters containing less than 500 parts per million of sulfate and less than 1000 p.p.m. of chloride may be determined within a limit of 0.1 p.p.m. when a 100-ml. sample is used.

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.

Efficient One-Pot Synthesis of Colloidal Zirconium Oxide Nanoparticles for High-Refractive-Index Nanocomposites.

PubMed

Liu, Chao; Hajagos, Tibor Jacob; Chen, Dustin; Chen, Yi; Kishpaugh, David; Pei, Qibing

2016-02-01

Zirconium oxide nanoparticles are promising candidates for optical engineering, photocatalysis, and high-κ dielectrics. However, reported synthetic methods for the colloidal zirconium oxide nanoparticles use unstable alkoxide precursors and have various other drawbacks, limiting their wide application. Here, we report a facile one-pot method for the synthesis of colloidally stable zirconium oxide nanoparticles. Using a simple solution of zirconium trifluoroacetate in oleylamine, highly stable zirconium oxide nanoparticles have been synthesized with high yield, following a proposed amidization-assisted sol-gel mechanism. The nanoparticles can be readily dispersed in nonpolar solvents, forming a long-term stable transparent solution, which can be further used to fabricate high-refractive-index nanocomposites in both monolith and thin-film forms. In addition, the same method has also been extended to the synthesis of titanium oxide nanoparticles, demonstrating its general applicability to all group IVB metal oxide nanoparticles.

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

Zirconium Phosphate Supported MOF Nanoplatelets.

PubMed

Kan, Yuwei; Clearfield, Abraham

2016-06-06

We report a rare example of the preparation of HKUST-1 metal-organic framework nanoplatelets through a step-by-step seeding procedure. Sodium ion exchanged zirconium phosphate, NaZrP, nanoplatelets were judiciously selected as support for layer-by-layer (LBL) assembly of Cu(II) and benzene-1,3,5-tricarboxylic acid (H3BTC) linkers. The first layer of Cu(II) is attached to the surface of zirconium phosphate through covalent interaction. The successive LBL growth of HKUST-1 film is then realized by soaking the NaZrP nanoplatelets in ethanolic solutions of cupric acetate and H3BTC, respectively. The amount of assembled HKUST-1 can be readily controlled by varying the number of growth cycles, which was characterized by powder X-ray diffraction and gas adsorption analyses. The successful construction of HKUST-1 on NaZrP was also supported by its catalytic performance for the oxidation of cyclohexene.

Zirconium carbide as an electrocatalyst for the chromous-chromic redox couple

NASA Technical Reports Server (NTRS)

Gahn, R. F.; Reid, M. A.; Yang, C. Y. (Inventor)

1981-01-01

Zirconium carbide is used as a catalyst in a REDOX cell for the oxidation of chromous ions to chromic ions and for the reduction of chromic ions to chromous ions. The zirconium carbide is coated on an inert electronically conductive electrode which is present in the anode fluid of the cell.

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