Artefacts in multimodal imaging of titanium, zirconium and binary titanium–zirconium alloy dental implants: an in vitro study

Science.gov (United States)

Schöllchen, Maximilian; Aarabi, Ghazal; Assaf, Alexandre T; Rendenbach, Carsten; Beck-Broichsitter, Benedicta; Semmusch, Jan; Sedlacik, Jan; Heiland, Max; Fiehler, Jens; Siemonsen, Susanne

2017-01-01

Objectives: To analyze and evaluate imaging artefacts induced by zirconium, titanium and titanium–zirconium alloy dental implants. Methods: Zirconium, titanium and titanium–zirconium alloy implants were embedded in gelatin and MRI, CT and CBCT were performed. Standard protocols were used for each modality. For MRI, line–distance profiles were plotted to quantify the accuracy of size determination. For CT and CBCT, six shells surrounding the implant were defined every 0.5 cm from the implant surface and histogram parameters were determined for each shell. Results: While titanium and titanium–zirconium alloy induced extensive signal voids in MRI owing to strong susceptibility, zirconium implants were clearly definable with only minor distortion artefacts. For titanium and titanium–zirconium alloy, the MR signal was attenuated up to 14.1 mm from the implant. In CT, titanium and titanium–zirconium alloy resulted in less streak artefacts in comparison with zirconium. In CBCT, titanium–zirconium alloy induced more severe artefacts than zirconium and titanium. Conclusions: MRI allows for an excellent image contrast and limited artefacts in patients with zirconium implants. CT and CBCT examinations are less affected by artefacts from titanium and titanium–zirconium alloy implants compared with MRI. The knowledge about differences of artefacts through different implant materials and image modalities might help support clinical decisions for the choice of implant material or imaging device in the clinical setting. PMID:27910719

Nanophase intermetallic FeAl obtained by sintering after mechanical alloying

Energy Technology Data Exchange (ETDEWEB)

D' Angelo, L., E-mail: luisa.dangelo@gmail.co [Departamento de Mecanica, UNEXPO, Luis Caballero Mejias, Charallave (Venezuela, Bolivarian Republic of); D' Onofrio, L. [Facultad de Ciencias, Dpto. Fisica, Universidad Central de Venezuela, Caracas (Venezuela, Bolivarian Republic of); Gonzalez, G., E-mail: gemagonz@ivic.v [Laboratorio de Materiales, Centro Tecnologico, Instituto Venezolano de Investigaciones Cientificas, Apdo. 21827, Caracas 1020A (Venezuela, Bolivarian Republic of)

2009-08-26

The preparation of bulk nanophase materials from nanocrystalline powders has been carried out by the application of sintering at high pressure. Fe-50 at.%Al system has been prepared by mechanical alloying for different milling periods from 1 to 50 h, using vials and balls of stainless steel and a ball-to-powder weight ratio (BPR) of 8:1 in a SPEX 8000 mill. Sintering of the 5 and 50 h milled powders was performed under high uniaxial pressure at 700 deg. C. The characterization of powders from each interval of milling was performed by X-ray diffraction, Moessbauer spectroscopy, scanning and transmission electron microscopy. After 5 h of milling formation of a nanocrystalline alpha-Fe(Al) solid solution that remains stable up to 50 h occurs. The grain size decreases to 7 nm after 50 h of milling. The sintering of the milled powders resulted in a nanophase-ordered FeAl alloys with a grain size of 16 nm. Grain growth during sintering was very small due to the effect of the high pressure applied.

Quantitative analysis of nickel in zirconium and zircaloy; Dosage du nickel dans le zirconium et dans le zircaloy

Energy Technology Data Exchange (ETDEWEB)

Rastoix, M [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1957-07-01

A rapid spectrophotometric has been developed for determination of 10 to 1000 ppm of Ni in zirconium and zircaloy using dimethylglyoxime. Iron, copper, tin and chromium, do not interfere at the concentration usually present in zirconium and its alloys. (author) [French] On determine colorimetriquenent 10 a 1000 ppm de Ni dans le zirconium et le zircaloy par photo colorimetrie a 440 m{mu} de la dimethylglyoxime nickelique. Le dosage est rapide. Le fer, le cuivre, l'etain, le chrome ne genent pas aux concentrations habituellement rencontrees dans le zirconium et ses alliages. (auteur)

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

Science.gov (United States)

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.

Radiochemical neutron activation analysis of zirconium and zirconium-niobium alloys

International Nuclear Information System (INIS)

Tashimova, F.A.; Sadikov, I.I.; Salimov, M.

2004-01-01

Full text: Zirconium and zirconium-niobium alloys are used on nuclear technology, as fuel cladding of nuclear reactors. Their nuclear-physical, mechanical and thermophysical properties are influenced them matrix and impurity composition, therefore determination of matrix and impurity content of these materials is a very important task. Neutron activation analysis is one from multielemental and high sensible techniques that are widely applied in analysis of high purity materials. Investigation of nuclear-physical characteristics of zirconium has shown that instrumental variant NAA is unusable for analysis due to high radioactivity of a matrix. Therefore it is necessary carrying out radiochemical separation of impurity radionuclides from matrix. Study of the literature datum have shown, that zirconium and niobium are very well extracted from muriatic solution with 5% tributyl phosphineoxide (TBPO) solution in toluene and 0,75 M solution of di-2-ethyl hexyl phosphoric acid (HDEHP) in cyclohexanone. Investigation of these elements extraction in these systems has shown that more effective and selective separation of matrix radionuclides is achieved in HDEHP-3M HCI system. This system is also extracted and hafnium, witch is an accompanying element of zirconium and its high content prevented determination of other impurity elements in sample. Therefore we used extraction system HDEHP-3M HCl for analysis of zirconium and zirconium-niobium alloys in chromatographic variant. By measurement of distribution profile of a matrix and of elution curve of determined elements is established, that for effective separation of impurity and matrix radionuclides there is enough chromatographic column with diameter 1 cm and height of a sorbent layer 7 cm, thus volume of elute, necessary for complete elution of determinate elements is 35-40 ml. On the basis of the carried out researches the technique of radiochemical NAA of high purity zirconium and zirconium-niobium alloy, which allows to

Zirconium alloy fuel cladding resistant to PCI crack propagation

International Nuclear Information System (INIS)

Boyle, R.F.; Foster, J.P.

1987-01-01

A nuclear fuel element is described cladding tube comprising: concentric tubular layers of zirconium base alloys; the concentric tubular layers including an inner layer and outer layer; the outer layer metallurgically bonded to the inner layer; the outer layer composed of a first zirconium base alloy characterized by excellent resistance to corrosion caused by exposure to high temperature and pressure aqueous environments; the inner layer composed of a second zirconium base alloy consisting of: about 0.2 to 0.6 wt.% tin, about 0.03 to 0.11 wt.% iron, less than about 0.02 wt.% chromium, up to about 350 ppm oxygen and the remainder being zirconium and incidental impurities, and the inner layer characterized by improved resistance to crack propagation under reactor operating conditions compared to the first zirconium alloy

Review of corrosion phenomena on zirconium alloys, niobium, titanium, inconel, stainless steel, and nickel plate under irradiation

International Nuclear Information System (INIS)

Johnson, A.B. Jr.

1975-01-01

The role of nuclear fluxes in corrosion processes was investigated in ATR, ETR, PRTR, and in Hanford production reactors. Major effort was directed to zirconium alloy corrosion parameter studies. Corrosion and hydriding results are reported as a function of oxygen concentration in the coolant, flux level, alloy composition, surface pretreatment, and metallurgical condition. Localized corrosion and hydriding at sites of bonding to dissimilar metals are described. Corrosion behavior on specimens transferred from oxygenated to low-oxygen coolants in ETR and ATR experiments is compared. Mechanism studies suggest that a depression in the corrosion of the Zr--2.5Nb alloy under irradiation is due to radiation-induced aging. The radiation-induced onset of transition on several alloys is in general a gradual process which nucleates locally, causing areas of oxide prosity which eventually encompass the surface. Examination of Zry-2 process tubes reveals that accelerated corrosion has occurred in low-oxygen coolants. Hydrogen contents are relatively low, but show some localized profiles. Gross hydriding has occurred on process tubes containing aluminum spacers, apparently by a galvanic charging mechanism. Titanium paralleled Zry-2 in corrosion behavior under irradiation. Niobium corrosion was variable, but did not appear to be strongly influenced by radiation. Corrosion rates on Inconel and stainless steel were only slightly higher in-flux than out-of-reactor. Corrosion rates on nickel-plated aluminum appeared to vary substantially with preexposure treatments, but the rates generally were accelerated compared to rates on unirradiated coupons. (59 references, 11 tables, 12 figs.)

PROCESS OF DISSOLVING ZIRCONIUM ALLOYS

Science.gov (United States)

Shor, R.S.; Vogler, S.

1958-01-21

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

Analysis of hafnium in zirconium alloys

International Nuclear Information System (INIS)

Kondo, Isao; Sakai, Fumiaki; Ohuchi, Yoshifusa; Nakamura, Hisashi

1977-01-01

It is required to analyse alloying components and impurity elements in the acceptance analysis of zirconium alloys as the material for fuel cladding tubes and pressure tubes for advanced thermal reactors. Because of extreme similarity in chemical properties between zirconium and hafnium, about 100 ppm of hafnium is usually contained in zirconium alloys. Zircaloy-2 alloy and 2.5% Nb-zirconium with the addition of hafnium had been prepared as in-house standard samples for rapid analysis. Study was made on fluorescent X-ray analysis and emission spectral analysis to establish the analytical method. By using these in-house standard samples, acceptance analysis was successfully carried out for the fuel cladding tubes for advanced thermal reactors. Sulfuric acid solution was prepared from JAERI-Z 1, 2 and 3, the standard sample for zircaloy-2 prepared by the Analytical Committee on Nuclear Fuel and Reactor Materials, JAERI, and zirconium oxide (Hf 1 ppm/Zr). Standard Hf solution was added to the sulfuric acid solution step by step, to make up a series of the standard oxide samples by the precipitation process. By the use of these standard samples, the development of the analytical method and joint analysis were made by the three-member analytical technique research group including PNC. The analytical precision for the fluorescent X-ray analysis was improved by attaching a metallic yttrium filter to the window of an X-ray tube so as to suppress the effect due to zirconium matrix. The variation factor of the joint analysis was about 10% to show good agreement, and the indication value was determined. (Kobatake, H.)

Quantitative analysis of nickel in zirconium and zircaloy

International Nuclear Information System (INIS)

Rastoix, M.

1957-01-01

A rapid spectrophotometric has been developed for determination of 10 to 1000 ppm of Ni in zirconium and zircaloy using dimethylglyoxime. Iron, copper, tin and chromium, do not interfere at the concentration usually present in zirconium and its alloys. (author) [fr

Electron microscopy of nuclear zirconium alloys

International Nuclear Information System (INIS)

Versaci, R.A.; Ipohorski, Miguel

1986-01-01

Transmission electron microscopy observations of the microstructure of zirconium alloys used in fuel sheaths of nuclear power reactors are reported. Specimens were observed after different thermal and mechanical treatment, similar to those actually used during fabrication of the sheaths. Electron micrographs and electron diffraction patterns of second phase particles present in zircaloy-2 and zircaloy-4 were also obtained, as well as some characteristic parameters. Images of oxides and hydrides most commonly present in zirconium alloys are also shown. Finally, the structure of a Zr-2,5Nb alloy used in CANDU reactors pressure tubes, is observed by electron microscopy. (Author) [es

Laves intermetallics in stainless steel-zirconium alloys

International Nuclear Information System (INIS)

Abraham, D.P.; McDeavitt, S.M.; Richardson, J.W. Jr.

1997-01-01

Laves intermetallics have a significant effect on properties of metal waste forms being developed at Argonne National Laboratory. These waste forms are stainless steel-zirconium alloys that will contain radioactive metal isotopes isolated from spent nuclear fuel by electrometallurgical treatment. The baseline waste form composition for stainless steel-clad fuels is stainless steel-15 wt.% zirconium (SS-15Zr). This article presents results of neutron diffraction measurements, heat-treatment studies and mechanical testing on SS-15Zr alloys. The Laves intermetallics in these alloys, labeled Zr(Fe,Cr,Ni) 2+x , have both C36 and C15 crystal structures. A fraction of these intermetallics transform into (Fe,Cr,Ni) 23 Zr 6 during high-temperature annealing; the authors have proposed a mechanism for this transformation. The SS-15Zr alloys show virtually no elongation in uniaxial tension, but exhibit good strength and ductility in compression tests. This article also presents neutron diffraction and microstructural data for a stainless steel-42 wt.% zirconium (SS-42Zr) alloy

Waterside corrosion of zirconium alloys in nuclear power plants

International Nuclear Information System (INIS)

1998-01-01

Technically the study of corrosion of zirconium alloys in nuclear power reactors is a very active field and both experimental work and understanding of the mechanisms involved are going through rapid changes. As a result, the lifetime of any publication in this area is short. Because of this it has been decided to revise IAEA-TECDOC-684 - Corrosion of Zirconium Alloys in Nuclear Power Plants - published in 1993. This updated, revised and enlarged version includes major changes to incorporate some of the comments received about the first version. Since this review deals exclusively with the corrosion of zirconium and zirconium based alloys in water, and another separate publication is planned to deal with the fuel-side corrosion of zirconium based fuel cladding alloys, i.e. stress corrosion cracking, it was decided to change the original title to Waterside Corrosion of Zirconium Alloys in Nuclear Power Plants. The rapid changes in the field have again necessitated a cut-off date for incorporating new data. This edition incorporates data up to the end of 1995; including results presented at the 11 International Symposium on Zirconium in the Nuclear Industry held in Garmisch-Partenkirchen, Germany, in September 1995. The revised format of the review now includes: Introductory chapters on basic zirconium metallurgy and oxidation theory; A revised chapter discussing the present extent of our knowledge of the corrosion mechanism based on laboratory experiments; a separate and revised chapter discussing hydrogen uptake; a completely reorganized chapter summarizing the phenomenological observations of zirconium alloy corrosion in reactors; a new chapter on modelling in-reactor corrosion; a revised chapter devoted exclusively to the manner in which irradiation might influence the corrosion process; finally, a summary of our present understanding of the corrosion mechanisms operating in reactor

Some recent trends in the use of zirconium alloys for nuclear service

International Nuclear Information System (INIS)

Balaramamoorthy, K.

1992-01-01

Without any exception nuclear power reactors particularly the water cooled ones, operating in the World use natural or slightly enriched uranium oxide fuel pellets with zirconium alloy cladding. While the zirconium alloys have proven to be successful in their designed usage, a desire for longer lifetimes of core components and increased duty cycle puts more demand on materials performance. This demand has led to more in depth studies of phenomena associated with zirconium alloy corrosion mechanism, fine tuning of the zirconium alloy composition, development of fabrication techniques and to the evaluation of newer zirconium alloys for critical applications. (author). 5 refs., 32 figs

Thermodynamic Database for Zirconium Alloys

International Nuclear Information System (INIS)

Jerlerud Perez, Rosa

2003-05-01

For many decades zirconium alloys have been commonly used in the nuclear power industry as fuel cladding material. Besides their good corrosion resistance and acceptable mechanical properties the main reason of using these alloys is the low neutron absorption. Zirconium alloys are exposed to a very severe environment during the nuclear fission process and there is a demand for better design of this material. To meet this requirement a thermodynamic database is developed to support material designers. In this thesis some aspects about the development of a thermodynamic database for zirconium alloys are presented. A thermodynamic database represents an important facility in applying thermodynamic equilibrium calculations for a given material providing: 1) relevant information about the thermodynamic properties of the alloys e.g. enthalpies, activities, heat capacity, and 2) significant information for the manufacturing process e.g. heat treatment temperature. The basic information in the database is first the unary data, i.e. pure elements; those are taken from the compilation of the Scientific Group Thermodata Europe (SGTE) and then the binary and ternary systems. All phases present in those binary and ternary systems are described by means of the Gibbs energy dependence on composition and temperature. Many of those binary systems have been taken from published or unpublished works and others have been assessed in the present work. All the calculations have been made using Thermo C alc software and the representation of the Gibbs energy obtained by applying Calphad technique

On the corrosion behaviour of stainless steel, nickel-chromium and zirconium-alloys in pore water of Portland cement

International Nuclear Information System (INIS)

Heitz, E.; Graefen, H.

1991-12-01

On the basis of an extensive review of literature and available experience, an evaluation was made of the corrosion of a metallic matrix for radioactive nuclides embedded in porous, water containing Portland cement. As a metallic matrix, austenitic high-alloy steel, nickel-base alloys and zirconium alloys are discussed. Pore waters in Portland cement have low aggressivity. However, through contact with formation water, chloride and sulphate enrichment can occur. Although corrosion is principally possible on the basis of purely thermodynamic considerations, it can be assumed that local corrosion (pitting, stress corrosion cracking, intergranular corrosion) is highly improbable under the given boundary conditions. This is valid for all three groups of alloys and means that only low release rates of corrosion products are to be expected. As a result of the discussion on radiolysis-induced corrosion, additional corrosion activity can be excluded. Final conclusions concerning the stimulation of corrosion processes by microbial action cannot be drawn and, therefore, additional experiments are proposed. The release rates of radioactive products are controlled by a very low dissolution rate of the materials in the passive state. All three groups of alloys show this type of general dissolution. From a survey of literature data it can be concluded that release rates greater than 250 mg/m 2 per day are not exceeded. Since these data were mainly obtained by electrochemical methods, it is proposed that quantitative analytical investigations of the corrosion products in pore water be made. On the whole the release rates determined are far below corrosion rates which are generally technically relevant. (author) 13 figs., 9 tabs., 61 refs

Superconductivity in zirconium-rhodium alloys

Science.gov (United States)

Zegler, S. T.

1969-01-01

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

A half-century of changes in zirconium alloys

International Nuclear Information System (INIS)

Mardon, J.P.; Barberis, P.; Hoffmann, P.B.

2008-01-01

This article presents the history of zirconium alloys for PWR and BWR technologies. For more than 20 years zirconium alloys have evolved to cope with demands of the reactor operators concerning the burn-up extension and new safety margins. The poor properties of Zircaloy-1 concerning corrosion have led researchers to add elements like iron by developing Zircaloy-3A and Zircaloy-3C, and resulting in Zircaloy-4 with tin addition (from 1.30% to 1.50%). Zircaloy-4 is now outdated for PWR and new zirconium alloys with niobium are used (M5, ZIRLO...) they present a better resistance to corrosion, to hydridation, to creep and they are less prone to dimensional changes under irradiation. (A.C.)

Manufacturing process to reduce large grain growth in zirconium alloys

International Nuclear Information System (INIS)

Rosecrans, P.M.

1987-01-01

A method is described of treating cold worked zirconium alloys to reduce large grain growth during thermal treatment above its recrystallization temperature. The method comprises heating the zirconium alloy at a temperature of about 1300 0 F. to 1350 0 F. for about 1 to 3 hours subsequent to cold working the zirconium alloy and prior to the thermal treatment at a temperature of between 1450 0 -1550 0 F., the thermal treatment temperature being above the recrystallization temperature

ZIRCONIUM-TITANIUM-BERYLLIUM BRAZING ALLOY

Science.gov (United States)

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

1962-06-12

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

Hydrogen content in titanium and a titanium–zirconium alloy after acid etching

Energy Technology Data Exchange (ETDEWEB)

Frank, Matthias J.; Walter, Martin S. [Department of Biomaterials, Institute for Clinical Dentistry, University of Oslo, P.O. Box 1109, Blindern, NO-0317 Oslo (Norway); Institute of Medical and Polymer Engineering, Chair of Medical Engineering, Technische Universität München, Boltzmannstrasse 15, 85748 Garching (Germany); Lyngstadaas, S. Petter [Department of Biomaterials, Institute for Clinical Dentistry, University of Oslo, P.O. Box 1109, Blindern, NO-0317 Oslo (Norway); Wintermantel, Erich [Institute of Medical and Polymer Engineering, Chair of Medical Engineering, Technische Universität München, Boltzmannstrasse 15, 85748 Garching (Germany); Haugen, Håvard J., E-mail: h.j.haugen@odont.uio.no [Department of Biomaterials, Institute for Clinical Dentistry, University of Oslo, P.O. Box 1109, Blindern, NO-0317 Oslo (Norway)

2013-04-01

Dental implant alloys made from titanium and zirconium are known for their high mechanical strength, fracture toughness and corrosion resistance in comparison with commercially pure titanium. The aim of the study was to investigate possible differences in the surface chemistry and/or surface topography of titanium and titanium–zirconium surfaces after sand blasting and acid etching. The two surfaces were compared by X-ray photoelectron spectroscopy, secondary ion mass spectroscopy, scanning electron microscopy and profilometry. The 1.9 times greater surface hydrogen concentration of titanium zirconium compared to titanium was found to be the major difference between the two materials. Zirconium appeared to enhance hydride formation on titanium alloys when etched in acid. Surface topography revealed significant differences on the micro and nanoscale. Surface roughness was increased significantly (p < 0.01) on the titanium–zirconium alloy. High-resolution images showed nanostructures only present on titanium zirconium. - Highlights: ► TiZr alloy showed increased hydrogen levels over Ti. ► The alloying element Zr appeared to catalyze hydrogen absorption in Ti. ► Surface roughness was significantly increased for the TiZr alloy over Ti. ► TiZr alloy revealed nanostructures not observed for Ti.

Techniques for chemical characterization of zirconium and its alloys

International Nuclear Information System (INIS)

Iyer, K.V.; Bassan, M.K.T.; Sudersanan, M.

2002-01-01

Chemical characterization of zirconium and its alloys such as zircaloy, Zr-Nb, etc for minor and trace constituents like Nb, Ti, Fe, Cr, Ni, Sn, Al etc has been carried out. Zirconium, being a major constituent, has been determined by gravimetry as zirconium oxide while other constituents like Nb, Ti, Fe have been determined by spectrophotometric methods. Other metals of importance at trace level have been estimated by AAS or ICPAES. The judicious use of both conventional and modern instrumental methods of analysis helps in the characterization of zirconium and its alloys for various major and minor constituents. The role of matrix effect in the determination was also investigated and methods have been worked out based on a preliminary separation of zirconium by a hydroxide precipitation. (author)

Development of tantalum–zirconium alloy for hydrogen purification

Energy Technology Data Exchange (ETDEWEB)

Kumar, Sanjay, E-mail: sanjay.barc@gmail.com [Fusion Reactor Materials Section, MG, BARC, Mumbai 85 (India); IAMR, Hiroshima University, Higashihiroshima 739-8530 (Japan); Singh, Anamika [GSASM Hiroshima University, Higashihiroshima 739-8530 (Japan); Jain, Uttam; Dey, Gautam Kumar [Fusion Reactor Materials Section, MG, BARC, Mumbai 85 (India)

2016-11-01

Highlights: • Terminal solid solubility of Ta increases with Zr addition. • Increase in lattice parameters of Ta due to Zr addition may be the possible reason. • Enhance H solubility could also be explained on the change in e-DOS of Ta–Zr alloys. • Ta–Zr alloys could be possible combination for hydrogen purification membrane. - Abstract: Terminal solid solubility of hydrogen in Ta–Zr alloys has been studied in connection with the development of tantalum based metallic membrane for hydrogen/tritium purification. The alloys were prepared by vacuum arc melting technique and subsequently cold rolled to 0.2 mm thickness. The terminal solid solubility of hydrogen in these cold rolled samples was investigated in a modified Sieverts apparatus. The terminal solid solubility of hydrogen was marginally increased with zirconium content. The change in the lattices parameter of tantalum upon zirconium addition and the higher affinity of zirconium for hydrogen as compared to tantalum could be the possible reasons.

Neutron activation of chlorine in zirconium and zirconium alloys use of the matrix as comparator

International Nuclear Information System (INIS)

Cohen, I.M.; Gomez, C.D.; Mila, M.I.

1981-01-01

A procedure is described for neutron activation analysis of chlorine in zirconium and zirconium alloys. Calculation of chlorine concentration is performed relative to zirconium concentration in the matrix in order to minimize effects of differences in irradiation and counting geometry. Principles of the method and the results obtained are discussed. (author)

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

Science.gov (United States)

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 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

Science.gov (United States)

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 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

Development of zirconium alloy tube manufacturing technology

International Nuclear Information System (INIS)

Kim, In Kyu; Park, Chan Hyun; Lee, Seung Hwan; Chung, Sun Kyo

2009-01-01

In late 2004, Korea Nuclear Fuel Company (KNF) launched a government funded joint development program with Westinghouse Electric Co. (WEC) to establish zirconium alloy tube manufacturing technology in Korea. Through this program, KNF and WEC have developed a state of the art facility to manufacture high quality nuclear tubes. KNF performed equipment qualification tests for each manufacturing machine with the support of WEC, and independently carried out product qualification tests for each tube product to be commercially produced. Apart from those tests, characterization test program consisting of specification test and characterization test was developed by KNF and WEC to demonstrate to customers of KNF the quality equivalency of products manufactured by KNF and WEC plants respectively. As part of establishment of performance evaluation technology for zirconium alloy tube in Korea, KNF carried out analyses of materials produced for the characterization test program using the most advanced techniques. Thanks to the accomplishment of the development of zirconium alloy tube manufacturing technology, KNF is expected to acquire positive spin off benefits in terms of technology and economy in the near future

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

Science.gov (United States)

Mariani, Robert Dominick

2014-09-09

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

Mechanical and irradiation properties of zirconium alloys irradiated in HANARO

International Nuclear Information System (INIS)

Kwon, Oh Hyun; Eom, Kyong Bo; Kim, Jae Ik; Suh, Jung Min; Jeon, Kyeong Lak

2011-01-01

These experimental studies are carried out to build a database for analyzing fuel performance in nuclear power plants. In particular, this study focuses on the mechanical and irradiation properties of three kinds of zirconium alloy (Alloy A, Alloy B and Alloy C) irradiated in the HANARO (High-flux Advanced Neutron Application Reactor), one of the leading multipurpose research reactors in the world. Yield strength and ultimate tensile strength were measured to determine the mechanical properties before and after irradiation, while irradiation growth was measured for the irradiation properties. The samples for irradiation testing are classified by texture. For the irradiation condition, all samples were wrapped into the capsule (07M-13N) and irradiated in the HANARO for about 100 days (E > 1.0 MeV, 1.1 10 21 n/cm 2 ). These tests and results indicate that the mechanical properties of zirconium alloys are similar whether unirradiated or irradiated. Alloy B has shown the highest yield strength and tensile strength properties compared to other alloys in irradiated condition. Even though each of the zirconium alloys has a different alloying content, this content does not seem to affect the mechanical properties under an unirradiated condition and low fluence. And all the alloys have shown the tendency to increase in yield strength and ultimate tensile strength. Transverse specimens of each of the zirconium alloys have a slightly lower irradiation growth tendency than longitudinal specimens. However, for clear analysis of texture effects, further testing under higher irradiation conditions is needed

Development of microstructure in thermomechanical processing of zirconium alloys

International Nuclear Information System (INIS)

Jha, S.K.; Saibaba, N.; Jayaraj, R.N.

2009-01-01

Zirconium based alloys are used for the manufacture of fuel tubes pressure tubes calandria tubes and other components of Pressurized Heavy Water Reactors (PHWRS). In single or two phase zirconium alloy system a variety of microstructure can be generated by suitable heat treatments by the process of equilibrium and non equilibrium phase transformations Microstructure can also be modified by alloying with α and β stabilizers. The microstructure in Zr alloys could be single hexagonal phase (α alloys) two phase bcc and hexagonal (α + β alloys) phase, single metastable martensitic microstructure and β with ω phase. The microstructural and micro textural evolution during thermo mechanical treatments depends strongly on such initial microstructure. Hot extrusion is a significant bulk deformation step which decides the initial microstructure of the alloy. It is carried out at elevated temperature i e above the recrystallization temperature, which enable imposition of large strains in single step. This deformation causes a significant change in the microstructure of the material and depends on extrusion process parameters such as temperature, strain rate (Ram speed), reduction ratio etc. In the present paper development of microstructures, microtexture and texture have been examined. An attempt is also made to optimise the hot working parameters for different Zirconium alloys with help of these studies. (author)

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

International Nuclear Information System (INIS)

Korenko, M.K.; Merrick, H.F.; Gibson, R.C.

1982-01-01

An iron-nickel-chromium age-hardenable alloy suitable for use in fast breeder reactor ducts and cladding utilizes the gamma-double prime strengthening phase and has a morphology of the gamma-double prime phase enveloping the gamma-prime phase and delta phase distributed at or near the grain boundaries. The alloy consists essentially of about 40-50 percent nickel, 7.5-14 percent chromium, 1.5-4 percent niobium, .25-.75 percent silicon, 1-3 percent titanium, .1-.5 percent aluminum, .02-1 percent carbon, .002-.015 percent boron, and the balance iron. Up to 2 percent manganese and up to .01 percent magnesium may be added to inhibit trace element effects; up to .1 percent zirconium may be added to increase radiation swelling resistance; and up to 3 percent molybdenum may be added to increase strength

Research on development and application of titanium and zirconium alloys

International Nuclear Information System (INIS)

Suzuki, Toshiyuki; Sasano, Hisaoki; Uehara, Shigeaki; Nakano, Osamu; Shibata, Michio

1983-01-01

It can be said that titanium and zirconium are new metals from the viewpoint of the history of metals, but both have grown to the materials supporting modern industries, titanium alloys in aerospace and ocean development, and zirconium alloys in nuclear power application. However, the properties of both alloys have not yet been clarified. In this study, the synthesis of TiNi and its properties, precipitation hardening type titanium alloys, and the effect of oxygen on the mechanical properties of both alloys were examined. TiNi is the typical intermetallic compound which shows the peculiar properties. The method of its synthesis by diffusion was examined, and it was clarified that it is useful as a structural material and also as a functional material. Precipitation hardening type alloys have not been developed in titanium alloys, but in this study, the feasibility of several alloy systems was found. Both titanium and zirconium have large affinity to oxygen, and the oxygen absorbed in the manufacturing process cannot be reduced. The tensile property of both alloys was examined in wide temperature range, and the effect of oxygen was clarified. (Kako, I.)

Chemical sensitive interfacial free volume studies of nanophase Al-rich alloys

International Nuclear Information System (INIS)

Lechner, W.; Puff, W.; Wuerschum, R.; Wilde, G.

2006-01-01

Full text: Al-based nanocrystalline alloys have attracted substantial interest due to their outstanding mechanical properties. These alloys can be obtained by crystallization of melt-spun amorphous precursors or by grain refinement upon repeated cold-rolling of elemental layers. For both synthesis routes, the nanocrystallization process is sensitively affected by interfacial chemistry and free volumes. In order to contribute to an atomistic understanding of the interfacial structure and processes during nanocrystallization, the present work deals with studies of interfacial free volumes by means of positron-annihilation-spectroscopy. In addition to positron lifetime spectroscopy which yields information on the size of free volumes, coincident Doppler broadening of the positron-electron annihilation photons is applied as novel technique for studying the chemistry of interfaces in nanophase materials on an atomistic scale. Al-rich alloys of the above mentioned synthesis routes were studied in this work. (author)

METHOD AND ALLOY FOR BONDING TO ZIRCONIUM

Science.gov (United States)

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

1960-04-19

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

Fluorimetric determination of uranium in zirconium and zircaloy alloys

International Nuclear Information System (INIS)

Acosta L, E.

1991-05-01

The objective of this procedure is to determine microquantities of uranium in zirconium and zircaloy alloys. The report also covers the determination of uranium in zirconium alloys and zircaloy in the range from 0.25 to 20 ppm on 1 g of base sample of radioactive material. These limit its can be variable if the size of the used aliquot one is changed for the final determination of uranium. (Author)

High strength corrosion-resistant zirconium aluminum alloys

International Nuclear Information System (INIS)

Schulson, E.M.; Cameron, D.J.

1976-01-01

A zirconium-aluminum alloy is described possessing superior corrosion resistance and mechanical properties. This alloy, preferably 7.5-9.5 wt% aluminum, is cast, worked in the Zr(Al)-Zr 2 Al region, and annealed to a substantially continuous matrix of Zr 3 Al. (E.C.B.)

AN ELECTROPLATING METHOD OF FORMING PLATINGS OF NICKEL, COBALT, NICKEL ALLOYS OR COBALT ALLOYS

DEFF Research Database (Denmark)

1997-01-01

An electroplating method of forming platings of nickel, cobalt, nickel alloys or cobalt alloys with reduced stresses in an electrodepositing bath of the type: Watt's bath, chloride bath or a combination thereof, by employing pulse plating with periodic reverse pulse and a sulfonated naphthalene...

Neutronographic Texture Analysis of Zirconium Based Alloys

International Nuclear Information System (INIS)

Kruz'elová, M; Vratislav, S; Kalvoda, L; Dlouhá, M

2012-01-01

Neutron diffraction is a very powerful tool in texture analysis of zirconium based alloys used in nuclear technique. Textures of five samples (two rolled sheets and three tubes) were investigated by using basal pole figures, inversion pole figures, and ODF distribution function. The texture measurement was performed at diffractometer KSN2 on the Laboratory of Neutron Diffraction, Department of Solid State Engineering, Faculty of Nuclear Sciences and Physical Engineering, CTU in Prague. Procedures for studying textures with thermal neutrons and procedures for obtaining texture parameters (direct and inverse pole figures, three dimensional orientation distribution function) are also described. Observed data were processed by software packages HEXAL and GSAS. Our results can be summarized as follows: i) All samples of zirconium alloys show the distribution of middle area into two maxima in basal pole figures. This is caused by alloying elements. A characteristic split of the basal pole maxima tilted from the normal direction toward the transverse direction can be observed for all samples, ii) Sheet samples prefer orientation of planes (100) and (110) perpendicular to rolling direction and orientation of planes (002) perpendicular to normal direction, iii) Basal planes of tubes are oriented parallel to tube axis, meanwhile (100) planes are oriented perpendicular to tube axis. Level of resulting texture and maxima position is different for tubes and for sheets. The obtained results are characteristic for zirconium based alloys.

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

Energy Technology Data Exchange (ETDEWEB)

Meier, R. [European Commission, Joint Research Centre (JRC), Institute for Transuranium Elements (ITU), Postfach 2340, Karlsruhe 76125 (Germany); Heidelberg University, Institute of Physical Chemistry, Im Neuenheimer Feld 253, Heidelberg 69120 (Germany); Souček, P., E-mail: Pavel.Soucek@ec.europa.eu [European Commission, Joint Research Centre (JRC), Institute for Transuranium Elements (ITU), Postfach 2340, Karlsruhe 76125 (Germany); Malmbeck, R.; Krachler, M.; Rodrigues, A.; Claux, B.; Glatz, J.-P. [European Commission, Joint Research Centre (JRC), Institute for Transuranium Elements (ITU), Postfach 2340, Karlsruhe 76125 (Germany); Fanghänel, Th. [European Commission, Joint Research Centre (JRC), Institute for Transuranium Elements (ITU), Postfach 2340, Karlsruhe 76125 (Germany); Heidelberg University, Institute of Physical Chemistry, Im Neuenheimer Feld 253, Heidelberg 69120 (Germany)

2016-04-15

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. - Highlights: • Recovery of actinides was shown by electrorefining of U/Pu–Zr alloys in LiCl–KCl. • Constant current density of 20 mA/cm{sup 2} is applied. • Most of the actinides were dissolved avoiding zirconium co-dissolution. • Deterioration of the deposit quality by a small amount of co-deposited Zr is not observed.

Superficial effects during the activation of zirconium AB2 alloys

International Nuclear Information System (INIS)

Zerbino, J; Visitin, A; Triaca, W

2005-01-01

The activation of zirconium nickel alloys with and without the addition of chromium and titanium is investigated through electrochemical and optical techniques.These alloys show high hydrogen absorption capacity and are extensively used in metal hydride batteries.Recent investigations in aqueous 1 M KOH indicate oxide layer growth and occlusion of hydrogen species in the alloys during the application of different cathodic potential programmes currently used in the activation process.In this research several techniques such as voltammetry, ellipsometry, energy dispersive analysis of X-rays EDAX, and scanning electron microscopy SEM are applied on the polished massive alloy Zr 1 -xTi x , x=0.36 y 0.43, and Zr 1 -xTi x CrNi, x=0.1,0.2 y 0.4.Data analysis shows that the stability, compactness and structure of the passive layers are strongly dependent on the applied potential programme.The alloy activation depends on the formation of deepen crevices that remain after a new polishing. Microscopic observation shows increase in the crevices thickness after the cathodic sweep potential cycling, which produces fragmentation of the grains and oxide growth during the activation process.This indicates metal breaking and intergranular dissolution that take place together with oxide and hydride formation.In some cases the resultant crevice thickness is one or two orders higher than that of the superficial oxide growth indicating intergranular localised corrosion

Low cycle fatigue behaviour of zirconium alloys at 3000C

International Nuclear Information System (INIS)

Hosbons, R.R.

1975-01-01

The low cycle fatigue lives of two zirconium alloys, zirconium--2.5 wt percent niobium and zirconium--1.1 wt percent chromium--0.1 wt percent iron, have been determined at 300 0 C. Both annealed material and cold-worked and stress-relieved material have similar fatigue lives to annealed Zircaloy-2 but β-quenched zirconium--niobium and zirconium--chromium--iron have lower fatigue lives than annealed Zircaloy-2. An atmosphere containing a concentration of iodine lower than that required for stress corrosion cracking still significantly lowers the fatigue life. A mathematical relationship between fatigue life and short-term tensile properties was used to estimate the fatigue life of zirconium alloy fuel sheaths and it was estimated that for a strain cycle of 0.1 percent a cyclic frequency exceeding 0.116 Hz (10,000 cycles/ day) would be required to cause fatigue failure of the sheath before its design life is realized

Low cycle fatigue behaviour of zirconium alloys at 3000C

International Nuclear Information System (INIS)

Hosbons, R.R.

1975-01-01

The low cycle fatigue lives of two zirconium alloys, zirconium-2.5 wt% niobium and zirconium-1.1 wt% chronium-0.1 wt% iron, have been determined at 300 0 C. Both annealed material and cold-worked and stress-relieved material have similar fatigue lives to annealed Zircaloy-2 but β-quenched zirconium-niobium and zirconium-chromium-iron have lower fatigue lives than annealed Zircaloy-2. An atmosphere containing a concentration of iodine lower than that required for stress corrosion cracking still significantly lowers the fatigue life. A mathematical relationship between fatigue life and short-term tensile properties was used to estimate the fatigue life of zirconium alloy fuel sheaths and it was estimated that for a strain cycle of 0.1 per cent a cyclic frequency exceeding 0.116 Hz (10 000 cycles/day) would be required to cause fatigue failure of the sheath before its design life is realized. (author)

Titanium and zirconium alloys

International Nuclear Information System (INIS)

Pinard Legry, G.

1994-01-01

Titanium and zirconium pure and base alloys are protected by an oxide film with anionic vacancies which gives a very good resistance to corrosion in oxidizing medium, in some ph ranges. Results of pitting and crevice corrosion are given for Cl - , Br - , I - ions concentration with temperature and ph dependence, also with oxygenated ions effect. (A.B.). 32 refs., 6 figs., 3 tabs

Effect of Bi on the corrosion resistance of zirconium alloys

International Nuclear Information System (INIS)

Yao Meiyi; Zhou Bangxin; Li Qiang; Zhang Weipeng; Zhu Li; Zou Linghong; Zhang Jinlong; Peng Jianchao

2014-01-01

In order to investigate systematically the effect of Bi addition on the corrosion resistance of zirconium alloys, different zirconium-based alloys, including Zr-4 (Zr-l.5Sn-0.2Fe-0.1Cr), S5 (Zr-0.8Sn-0.35Nb-0.4Fe-0.1Cr), T5 (Zr-0.7Sn-l.0Nb-0.3Fe-0.1Cr) and Zr-1Nb, were adopted to prepare the zirconium alloys containing Bi of 0∼0.5% in mass fraction. These alloys were denoted as Zr-4 + xBi, S5 + xBi, T5 + xBi and Zr-1Nb + xBi, respectively. The corrosion behavior of these specimens was investigated by autoclave testing in lithiated water with 0.01 M LiOH or deionized water at 360 ℃/18.6 MPa and in superheated steam at 400 ℃/10.3 MPa. The microstructure of the alloys was examined by TEM and the second phase particles (SPPs) were analyzed by EDS. Microstructure observation shows that the addition of Bi promotes the precipitation of Sn as second phase particles (SPPs) because Sn is in solid solution in α-Zr matrix in Zr-4, S5 and T5 alloys. The concentration of Bi dissolved in α-Zr matrix increase with the increase of Nb in the alloys, and the excess Bi precipitates as Bi-containing SPPs. The corrosion results show that the effect of Bi addition on the corrosion behavior of different zirconium-based alloys is very complicated, depending on their compositions and corrosion conditions. In the case of higher Bi concentration in α-Zr, the zirconium alloys exhibit better corrosion resistance. However, in the case of precipitation of Bi-containing SPPs, the corrosion resistance gets worse. This indicates that the solid solution of Bi in α-Zr matrix can improve the corrosion resistance, while the precipitation of the Bi-containing SPPs is harmful to the corrosion resistance. (authors)

Nickel base alloys

International Nuclear Information System (INIS)

Gibson, R.C.; Korenko, M.K.

1980-01-01

Nickel based alloy, the characteristic of which is that it mainly includes in percentages by weight: 57-63 Ni, 7-18 Cr, 10-20 Fe, 4-6 Mo, 1-2 Nb, 0.2-0.8 Si, 0.01-0.05 Zr, 1.0-2.5 Ti, 1.0-2.5 Al, 0.02-0.06 C and 0.002-0.015 B. The aim is to create new nickel-chromium alloys, hardened in a solid solution and by precipitation, that are stable, exhibit reduced swelling and resistant to plastic deformation inside the reactor. These alloys of the gamma prime type have improved mechanical strengthm swelling resistance, structural stability and welding properties compared with Inconel 625 [fr

Nickel alloys and high-alloyed special stainless steels. Properties, manufacturing, applications. 4. compl. rev. ed.

International Nuclear Information System (INIS)

Heubner, Ulrich; Kloewer, Jutta; Alves, Helena; Behrens, Rainer; Schindler, Claudius; Wahl, Volker; Wolf, Martin

2012-01-01

This book contains the following eight topics: 1. Nickel alloys and high-alloy special stainless steels - Material overview and metallurgical principles (U. Heubner); 2. Corrosion resistance of nickel alloys and high-alloy special stainless steels (U. Heubner); 3. Welding of nickel alloys and high-alloy special stainless steels (T. Hoffmann, M. Wolf); 4. High-temperature materials for industrial plant construction (J. Kloewer); 5. Nickel alloys and high-alloy special stainless steels as hot roll clad composites-a cost-effective alternative (C. Schindler); 6. Selected examples of the use of nickel alloys and high-alloy special stainless steels in chemical plants (H. Alves); 7. The use of nickel alloys and stainless steels in environmental engineering (V. Wahl); 8: Nickel alloys and high-alloy special stainless steels for the oil and gas industry (R. Behrens).

Microstructural characterization of mechanically alloyed Al–Cu–Mn alloy with zirconium

Energy Technology Data Exchange (ETDEWEB)

Prosviryakov, A.S., E-mail: pro.alex@mail.ru; Shcherbachev, K.D.; Tabachkova, N.Yu.

2015-01-19

An evolution of Al–Cu–Mn alloy microstructure during its mechanical alloying with zirconium 20 wt% and after subsequent annealing was studied by X-ray diffraction, light microscopy and transmission electron microscopy. The effect of milling time on powder microhardness, Al lattice parameter, lattice microstrain and crystallite size was determined.

Annealing texture of rolled nickel alloys

International Nuclear Information System (INIS)

Meshchaninov, I.V.; Khayutin, S.G.

1976-01-01

A texture of pure nickel and binary alloys after the 95% rolling and annealing has been studied. Insoluble additives (Mg, Zr) slacken the cubic texture in nickel and neral slackening of the texture (Zr). In the case of alloying with silicium (up to 2%) the texture practically coinsides with that of a technical-grade nickel. The remaining soluble additives either do not change the texture of pure nickel (C, Nb) or enhance the sharpness and intensity of the cubic compontnt (Al, Cu, Mn, Cr, Mo, W, Co -at their content 0.5 to 2.0%). A model is proposed by which variation of the annealing texture upon alloying is caused by dissimilar effect of the alloying elements on the mobility of high- and low-angle grain boundaries

Phase Transformations in a Uranium-Zirconium Alloy containing 2 weight per cent Zirconium

Energy Technology Data Exchange (ETDEWEB)

Lagerberg, G

1961-04-15

The phase transformations in a uranium-zirconium alloy containing 2 weight percent zirconium have been examined metallographically after heat treatments involving isothermal transformation of y and cooling from the -y-range at different rates. Transformations on heating and cooling have also been studied in uranium-zirconium alloys with 0.5, 2 and 5 weight per cent zirconium by means of differential thermal analysis. The results are compatible with the phase diagram given by Howlett and Knapton. On quenching from the {gamma}-range the {gamma} phase transforms martensitically to supersaturated a the M{sub S} temperature being about 490 C. During isothermal transformation of {gamma} in the temperature range 735 to 700 C {beta}-phase is precipitated as Widmanstaetten plates and the equilibrium structure consists of {beta} and {gamma}{sub 1}. Below 700 C {gamma} transforms completely to Widmanstaetten plates which consist of {beta} above 660 C and of a at lower temperatures. Secondary phases, {gamma}{sub 2} above 610 C and {delta} below this temperature, are precipitated from the initially supersaturated Widmanstaetten plates during the isothermal treatments. At and slightly below 700 C the cooperative growth of |3 and {gamma}{sub 2} is observed. The results of isothermal transformation are summarized in a TTTdiagram.

Recrystallization resistance in aluminum alloys containing zirconium

International Nuclear Information System (INIS)

Ranganathan, K.

1991-01-01

Zirconium forms a fine dispersion of the metastable β' (Al 3 Zr) phase that controls recrystallization by retarding the motion of high-angle boundaries. The primary material chosen for this research was aluminum alloy 7150 containing zinc, magnesium, and copper as the major solute elements and zirconium as the dispersoid-forming element. The size, distribution, and the volume fraction of β' was controlled by varying the alloy composition and preheat practices. Preheated ingots were subjected to a specific sequence of hot-rolling operations to evaluate the resistance to recrystallization of the different microstructures. Optical and transmission electron microscopy (TEM) techniques were used to investigate the influence of dispersoid morphology resulting from the thermal treatments and deformation processing on the recrystallization behavior of the alloy. Studies were conducted to determine the influence of the individual solute elements present in 7150 on the precipitation of β' and consequently on the recrystallization behavior of the material. These studies were done on compositional variants of commercial 7150

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

Science.gov (United States)

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)

Effect of homogenization heat treatments on the cast structure and tensile properties of nickel-base superalloy ATI 718Plus in the presence of boron and zirconium additions

Energy Technology Data Exchange (ETDEWEB)

Hosseini, Seyed Ali, E-mail: saliho3ini@gmail.com; Madar, Karim Zangeneh; Abbasi, Seyed Mehdi

2017-03-24

The effect of homogenization heat treatment on cast structure, hardness, and tensile properties of the nickel-based superalloy 718plus in the presence of boron and zirconium additives were investigated. For this purpose, five alloys with different contents of boron (0.00–0.016 wt%) and zirconium (0.0–0.1 wt%) were cast by double vacuum process VIM/VAR and then were homogenized at 1075–1175 °C for 5–25 h. Microstructural investigation by OM and SEM and phase analysis by XRD were done and then hardness and high temperature tensile tests were performed on the homogenized alloys. The results show that the amount of the Laves phase is reduced by increases in time and temperature of homogenization. It was also found that increases in duration of homogenization at 1075 °C results in improving strength and ductility, while duration increase at 1175 °C is accompanied with degradation of them, which caused the reduction of needle-like delta phase on grain boundaries. Boron and zirconium had negative effects on the strength and ductility of the alloy by increasing the amount of Laves in the cast structure. By increasing these elements in alloy composition, more time is needed in order to fully eliminate the Laves by homogenization treatment.

Development of Zirconium alloys (for pressure tubes)

Energy Technology Data Exchange (ETDEWEB)

Kim, Young Suk; Kwon, Sang Chul; Choo, Ki Nam; Jung, Chung Hwan; Yim, Kyong Soo; Kim, Sung Soo; Baek, Jong Hyuk; Jeong, Yong Hwan; Kim, Kyong Ho; Cho, Hae Dong [Korea Atomic Energy Research Inst., Daeduk (Korea, Republic of); Hwang, S. K.; Kim, M. H. [Inha Univ., Incheon (Korea, Republic of); Kwon, S. I [Korea Univ., Seoul (Korea, Republic of); Kim, I. S. [Korea Advanced Inst. of Science and Technology, Taejon (Korea, Republic of)

1997-09-01

The objective of this research is to set up the basic technologies for the evaluation of pressure tube integrity and to develop improved zirconium alloys to prevent pressure tube failures due to DHC and hydride blister caused by excessive creep-down of pressure tubes. The experimental procedure and facilities for characterization of pressure tubes were developed. The basic research related to a better understanding of the in-reactor performances of pressure tubes leads to noticeable findings for the first time : the microstructural effect on corrosion and hydrogen pick-up behavior of Zr-2.5Nb pressure tubes, texture effect on strength and DHC resistance and enhanced recrystallization by Fe in zirconium alloys and etc. Analytical methodology for the assessment of pressure tubes with surface flaws was set up. A joint research is being under way with AECL to determine the fracture toughness of O-8 at the EOL (End of Life) that had been quadruple melted and was taken out of the Wolsung Unit-1 after 10 year operation. In addition, pressure tube with texture controlled is being made along with VNINM in Russia as a joint project between KAERI and Russia. Finally, we succeeded in developing 4 different kinds of zirconium alloys with better corrosion resistance, low hydrogen pickup fraction and higher creep strength. (author). 121 refs., 65 tabs., 260 figs

Electrodeposition of zinc--nickel alloys coatings

Energy Technology Data Exchange (ETDEWEB)

Dini, J W; Johnson, H R

1977-10-01

One possible substitute for cadmium in some applications is a zinc--nickel alloy deposit. Previous work by others showed that electrodeposited zinc--nickel coatings containing about 85 percent zinc and 15 percent nickel provided noticeably better corrosion resistance than pure zinc. Present work which supports this finding also shows that the corrosion resistance of the alloy deposit compares favorably with cadmium.

Liquid phase sintering of carbides using a nickel-molybdenum alloy

International Nuclear Information System (INIS)

Barranco, J.M.; Warenchak, R.A.

1987-01-01

Liquid phase vacuum sintering was used to densify four carbide groups. These were titanium carbide, tungsten carbide, vanadium carbide, and zirconium carbide. The liquid phase consisted of nickel with additions of molybdenum of from 6.25 to 50.0 weight percent at doubling increments. The liquid phase or binder comprised 10, 20, and 40 percent by weight of the pressed powders. The specimens were tested using 3 point bending. Tungsten carbide showed the greatest improvement in bend rupture strength, flexural modulus, fracture energy and hardness using 20 percent binder with lesser amounts of molybdenum (6.25 or 12.5 wt %) added to nickel compared to pure nickel. A refinement in the carbide microstructure and/or a reduction in porosity was seen for both the titanium and tungsten carbides when the alloy binder was used compared to using the nickel alone. Curves depicting the above properties are shown for increasing amounts of molybdenum in nickel for each carbide examined. Loss of binder phase due to evaporation was experienced during heating in vacuum at sintering temperatures. In an effort to reduce porosity, identical specimens were HIP processed at 15 ksi and temperatures averaging 110 C below the sintering g temperature. The tungsten carbide and titanium carbide series containing 80 and 90 weight percent carbide phase respectively showed improvement properties after HIP while properties decreased for most other compositions

Corrosion resistant zirconium alloys prepared by powder metallurgy

International Nuclear Information System (INIS)

Wojeik, C.C.

1984-01-01

Pure zirconium and zirconium 2.5% niobium were prepared by powder metallurgy. The powders were prepared directly from sponge and consolidated by cold isostatic pressing and sintering. Hot isostatic pressing was also used to obtain full density after sintering. For pure zirconium the effects of particle size, compaction pressure, sintering temperature and purity were investigated. Fully densified zirconium and Zr-2.5%Nb exhibited tensile properties comparable to cast material at room temperature and 300 0 F (149 0 C). Pressed and sintered material having density of 94-99% had slightly lower tensile properties. Corrosion tests were performed in boiling 65% H/sub 2/SO/sub 4/, 70% HNO/sub 3/, 20% HCl and 20% HCl + 500 ppm FeCl/sub 3/ (a known pitting solution). For fully dense material the observed corrosion behavior was nearly equivalent to cast material. A slightly higher rate of attack was observed for samples which were only 94-99% dense. Welding tests were also performed on zirconium and Zr-2.5%Nb alloy. Unlike P/M titanium alloys, these materials had good weldability due to the lower content of volatile impurities in the powder. A slight amount of weld porosity was observed but joint efficiencies were always not 100%, even for 94-99% density samples. Several practical applications of the P/M processed material will be briefly described

Towards an understanding of zirconium alloy corrosion

International Nuclear Information System (INIS)

Cox, B.

1976-08-01

A brief historical summary is given of the development of a programme for understanding the corrosion mechanisms operating for zirconium alloys. A general summary is given of the progress made, so far, in carrying through this programme. (author)

High temperature cathodic charging of hydrogen in zirconium alloys and iron and nickel base alloys

International Nuclear Information System (INIS)

John, J.T.; De, P.K.; Gadiyar, H.S.

1990-01-01

These investigations lead to the development of a new technique for charging hydrogen into metals and alloys. In this technique a mixture of sulfates and bisulfates of sodium and potassium is kept saturated with water at 250-300degC in an open pyrex glass beaker and electrolysed using platinum anode and the material to be charged as the cathode. Most of the studies were carried out on Zr alloys. It is shown that because of the high hydrogen flux available at the surface and the high diffusivity of hydrogen in metals at these temperatures the materials pick up hydrogen faster and more uniformly than the conventional electrolytic charging at room temperature and high temperature autoclaving in LiOH solutions. Chemical analysis, metallographic examination and XRD studies confirm this. This technique has been used to charge hydrogen into many iron and nickel base austentic alloys, which are very resistant to hydrogen pick up and to H-embrittlement. Since this involved a novel method of electrolysing water, the hydrogen/deuterium isotopic ratio has been studied. At this temperatures the D/H ratio in the evolved hydrogen gas was found to be closer to the value in the liquid water, which means a smaller separation factor. This confirm the earlier observation that separation factor decreases with increase of temperature. (author). 16 refs., 21 fi gs., 6 tabs

Control of microstructure during hot working of zirconium alloys

International Nuclear Information System (INIS)

Chakravartty, J.K.; Banerjee, S.

2005-01-01

Hot working is considered to be the most important step involved in the fabrication of zirconium alloys for nuclear reactor applications for two reasons: i) the scale of the microstructure and texture of the final product is decided at this stage and ii) the hot deformed microstructure provides a suitable starting microstructure for the subsequent fabrication steps. The resultant microstructure in turn controls the properties of the final product. In order to obtain final product with a suitable microstructure and with specified mechanical properties on a repeatable basis the control of microstructure during hot working is of paramount importance. This is usually done by studying the constitutive behaviour of the material under hot working conditions and by constructing processing maps. In the latter method, strain rate sensitivity is mapped as a function of temperature and strain rate to delineate domains within the bounds of which a specific deformation mechanism dominates. Detail microstructural analysis is then carried out on the samples deformed within the domains. Using this methodology, processing maps have been constructed for various zirconium alloys. These maps have been found to be very useful for optimizing the hot workability and control of microstructure of zirconium alloys. (author)

Thermo-mechanical treatment of zirconium alloys

International Nuclear Information System (INIS)

Levy, I.S.

1975-01-01

A zirconium alloy comprising at least 95 percent Zr (Zircaloy), which has been thoroughly annealed, is greatly increased in strength without substantial loss in ductility by subjecting it to tensile creep deformation in a temperature range in which creep will occur, yet which is below the temperature for significant recovery. (U.S.)

Cathodic protection of steel by electrodeposited zinc-nickel alloy coatings

Energy Technology Data Exchange (ETDEWEB)

Baldwin, K.R.; Smith, C.J.E. [Defence Research Agency, Farnborough (United Kingdom). Structural Materials Centre; Robinson, M.J. [Cranfield Univ. (United Kingdom). School of Industrial and Manufacturing Science

1995-12-01

The ability of electrodeposited zinc-nickel alloy coatings to cathodically protect steel was studied in dilute chloride solutions. The potential distribution along steel strips partly electroplated with zinc-nickel alloys was determined, and the length of exposed steel that was held below the minimum protection potential (E{sub prot}) was taken as a measure of the level of cathodic protection (CP) provided by the alloy coatings. The level of CP afforded by zinc alloy coatings was found to decrease with increasing nickel content. When nickel content was increased to {approx} {ge} 21 wt%, no CP was obtained. Surface analysis of uncoupled zinc-nickel alloys that were immersed in sodium chloride (NaCl) solutions showed the concentration of zinc decreased in the surface layers while the concentration of nickel increased, indicating that the alloys were susceptible to dezincification. The analysis of zinc-nickel alloy coatings on partly electroplated steel strips that were immersed in chloride solution showed a significantly higher level of dezincification than that found for uncoupled alloy coatings. This effect accounted for the rapid loss of CP afforded to steel by some zinc alloy coatings, particularly those with high initial nickel levels.

Deformation mechanisms and irradiation effects in zirconium alloys. A multi-scale study

International Nuclear Information System (INIS)

Onimus, Fabien

2015-01-01

Zirconium alloys have been used for more than 30 years in the nuclear industry as structural materials for the fuel assemblies of pressurized water reactors. In particular, the cladding tube, made of zirconium alloys, constitutes the first barrier against the dissemination of radioactive elements. It is therefore essential to have a good understanding and prediction of the mechanical behavior of these materials in various conditions. The work presented in this dissertation deals with an experimental study and numerical simulations, at several length scales, of the deformation mechanisms and the mechanical behavior of zirconium alloys before irradiation, but also after irradiation and under irradiation. The mechanical behavior of zirconium single crystal has been determined, during an original study, using tensile test specimens containing large grains. Based on this study, crystal plasticity constitutive laws have been proposed. A polycrystalline model has also been developed to simulate the behavior of unirradiated zirconium alloys. A thorough Transmission Electron Microscopy (TEM) study has been able to clarify the deformation mechanisms of zirconium alloys occurring after irradiation. The clearing of loops by gliding dislocations leading to the dislocation channeling mechanism has been studied in details. This phenomenon has also been simulated using a dislocation dynamics code. The macroscopic consequences of this process have also been analyzed. A polycrystalline model taking into account the specificity of this mechanism has eventually been proposed. This approach has then been extended to the post-irradiation creep behavior. The recovery of radiation defects during creep tests has been characterized by TEM and modeled using cluster dynamics method. Deformation modes during creep have also been studied and a simple model for the creep behavior has eventually been proposed. Finally, the mechanism responsible for the acceleration of irradiation growth that

Surface coating Zr or Zr alloy nuclear fuel elements

International Nuclear Information System (INIS)

Donaghy, R.E.; Sherman, A.H.

1980-01-01

A method is disclosed for preventing stress corrosion cracking or metal embrittlement of a zirconium or zirconium alloy container that is to be coated on the inside surface with a layer of a metal such as copper, a copper alloy, nickel, or iron and used for holding nuclear fuel material as a nuclear fuel element. The zirconium material is etched in an etchant solution, desmutted mechanically or ultrasonically, oxidized to form an oxide coating on the zirconium, cleaned in an aqueous alkaline cleaning solution, activated for electroless deposition of a metal layer and contacted with an electroless metal plating solution. This method provides a boundary layer of zirconium oxide between the zirconium container and the metal layer. (author)

Diffusion of alloying elements in liquid nickel

International Nuclear Information System (INIS)

Ershov, G.S.; Majboroda, V.P.; Permyakova, T.V.

1990-01-01

Values of diffusion coefficients for chromium, vanadium, zinc, silicon, tin, antimony, lead and zirconium in liquid nickel are determined within 1500-1700 deg C temperature range using annular gap technique. The data obtained are explained concerning microheterogeneous structure of metallic melts

Stress corrosion crack tip microstructure in nickel-based alloys

International Nuclear Information System (INIS)

Shei, S.A.; Yang, W.J.

1994-04-01

Stress corrosion cracking behavior of several nickel-base alloys in high temperature caustic environments has been evaluated. The crack tip and fracture surfaces were examined using Auger/ESCA and Analytical Electron Microscopy (AEM) to determine the near crack tip microstructure and microchemistry. Results showed formation of chromium-rich oxides at or near the crack tip and nickel-rich de-alloying layers away from the crack tip. The stress corrosion resistance of different nickel-base alloys in caustic may be explained by the preferential oxidation and dissolution of different alloying elements at the crack tip. Alloy 600 (UNS N06600) shows good general corrosion and intergranular attack resistance in caustic because of its high nickel content. Thermally treated Alloy 690 (UNS N06690) and Alloy 600 provide good stress corrosion cracking resistance because of high chromium contents along grain boundaries. Alloy 625 (UNS N06625) does not show as good stress corrosion cracking resistance as Alloy 690 or Alloy 600 because of its high molybdenum content

Characterization of zirconium alloy oxidation films by alternating current impedance

International Nuclear Information System (INIS)

Rosecrans, P.M.

1984-01-01

Kinetics of zirconium alloy oxidation are highly nonlinear. The results of electrochemical measurements and electron microscopy support the existence of porosity in oxide films formed on zirconium alloys in high temperature aqueous environments. Analytical treatment is presented relating oxidation kinetics to the thickness and distribution of nonporous elements within the oxide. This analysis illustrates that both the level and distribution of porosity within the oxide factor into oxidation kinetics. The barrier layer model can provide a basis for predicting the effect of environmental changes on oxidation rate. In addition, it demonstrates the need for further research into porosity generation mechanisms in oxide films

In situ Investigation of Oxide Films on Zirconium Alloy in PWR Primary Water Chemistry

Energy Technology Data Exchange (ETDEWEB)

Kim, Taeho; Choi, Kyoung Joon; Yoo, Seung Chang; Kim, Ji Hyun [Ulsan National Institute of Science and Technology, Ulsan (Korea, Republic of)

2015-05-15

Zirconium alloys are used as fuel cladding materials in nuclear power reactors, because these materials have a very low thermal neutron capture cross section as well as desirable mechanical properties. However, the Fukushima accident shows that the oxidation behavior of zirconium alloy is an important issue because the zirconium alloy functions as a shield of nuclear material (i.e., uranium, fission gas), and the degradation on zirconium cladding directly causes severe accident on nuclear power plant. Therefore, to ensure the safety of nuclear power reactors, the performance and sustainability of nuclear fuel should be understood. Currently, the water-metal interface is regarded as the rate-controlling site governing the rapid oxidation transition in high-burn-up fuels. Zirconium oxide is formed at the water-metal interface, and its structure and phase play an important role in determining its mechanical properties. In the early stage of the oxidation process, zirconium oxide with both tetragonal and monoclinic phases is formed. With an increase in the oxidation time to 150 h, the unstable tetragonal phase disappears and the monoclinic phase is dominant and possibly because of the stress relaxation according to previous and present results.

Round robin test for zirconium alloys in 400 deg C steam: results from EDF

International Nuclear Information System (INIS)

Blat, M.

1994-01-01

The EDF Material Studies Branch has participated in the Round Robin program of uniform corrosion on zirconium alloys. The objectives of these Round Robin corrosion tests are to generate new uniform corrosion weight gain date utilizing modern zirconium alloy products and to improve the International and ASTM standards. (author). 2 tabs., 7 appendix., 2 refs

Long-time corrosion and high-temperature oxidation of zirconium alloys applied on NPP like fuel elements cover

International Nuclear Information System (INIS)

Vrtilkova, V.; Novotny, L.; Lingart, S.; Doukha, R.; Yarosh, Ya.; Kolenchik, Ya.

2007-01-01

Zirconium is applying in nuclear energy since 50-th of last century in capacity of material for cover production for fuel elements, reactor fuel and structural parts, and mainly due to both corrosion stability and low effective cross section for thermal neutrons capture. Impurities in doping elements form and alloy production technology has influence on mechanical and corrosion properties of finite alloy. Long-time corrosion tests for several zirconium alloys in forcing autoclave under different reaction conditions were carried out. After that process kinetics was studied, mass increase, hydrogen formation, zirconium hydride forming morphology, zirconium oxide layer thickness have been determined as well

Sliding wear and friction behavior of zirconium alloy with heat-treated Inconel718

Energy Technology Data Exchange (ETDEWEB)

Kim, J.H., E-mail: kimjhoon@cnu.ac.kr [Dept. of Mechanical Design Engineering, Chungnam National University, 99 Daehak-ro, Yuseong-gu, Daejeon 305-764 (Korea, Republic of); Park, J.M. [Dept. of Mechanical Design Engineering, Chungnam National University, 99 Daehak-ro, Yuseong-gu, Daejeon 305-764 (Korea, Republic of); Park, J.K.; Jeon, K.L. [Nuclear Fuel Technology Department, Korea Nuclear Fuel, 1047 Daedukdae-ro, Yuseong-gu, Daejeon 305-353 (Korea, Republic of)

2014-04-01

In water-cooled nuclear reactors, the sliding of fuel rod can lead to severe wear and it is an important issue to sustain the structural integrity of nuclear reactor. In the present study, sliding wear behavior of zirconium alloy in dry and water environment using Pin-On-Disk sliding wear tester was investigated. Wear resistance of zirconium alloy against heat-treated Inconel718 pin was examined at room temperature. Sliding wear tests were carried out at different sliding distance, axial load and sliding speed based on ASTM (G99-05). The results of these experiments were verified with specific wear rate and coefficient of friction. The micro-mechanisms responsible for wear in zirconium alloy were identified to be microcutting and microcracking in dry environment. Moreover, micropitting and delamination were observed in water environment.

Prospects for zirconium structural alloys at high temperatures

International Nuclear Information System (INIS)

Thomas, W.R.

1969-05-01

Improved station efficiencies and lower capital costs provide incentives for the development of zirconium alloys for pressure tubes which can operate at temperatures above 450 o C. The experience of the Ti industry indicates that a complex alloy containing solution hardeners of Sn or Al and precipitation hardeners of Mo and Nb and perhaps Si will be required. The thermal neutron cross-section of the alloy will be about 10% higher than Zircaloy-2 and because of its poor corrosion resistance will require cladding with a corrosion resistant alloy such as Zr-Cr. Results to date indicate that such a pressure tube is feasible. (author)

Design basis for creep of zirconium alloy components in a fast neutron flux

International Nuclear Information System (INIS)

Ross-Ross, P.A.; Fidleris, V.

1975-01-01

The chalk River Nuclear Laboratory's experience with the creep of zirconium alloys in a neutron flux is described. Fast neutron flux changes the creep behaviour of zirconium alloys and new design criteria for in-reactor applications are needed. From experimental results empirical relations describing the effects of neutron flux, stress, temperature, time and anisotropy on creep rate were established. The relations are applied to the design of pressure tubes. (author)

Design basis for creep of zirconium alloy components in a fast neutron flux

International Nuclear Information System (INIS)

Ross-Ross, P.A.; Fidleris, V.

1974-01-01

The Chalk River Nuclear Laboratory's experience with the creep of zirconium alloys in a neutron flux is described. Fast neutron flux changes the creep behavior of zirconium alloys and new design criteria for in-reactor applications are needed. From experimental results empirical relations describing the effects of neutron flux, stress, temperature, time, and anisotropy on creep rate were established. The relations are applied to the design of pressure tubes. (author)

Protection of zirconium and its alloys by metallic coatings

International Nuclear Information System (INIS)

Loriers, H.; Lafon, A.; Darras, R.; Baque, P.

1968-01-01

At 600 deg. C in an atmosphere of carbon dioxide, zirconium and its alloys undergo corrosion which presents two aspects simultaneously: - formation of a surface layer of zirconia, - dissolution of oxygen in the alloy sub-layer leading to brittleness. The two phenomena greatly restrict the possibilities of using zirconium alloys as a canning material for fuel elements in CO 2 cooled nuclear reactors. An attempt has thus been made to limit, and perhaps to suppress, the corrosion effects in zirconium under these conditions by protecting it with metallic coatings. A first attempt to obtain a protection using copper-based coatings did not produce the result hoped for. Aluminium coatings produced by vacuum evaporation, followed by a consolidating thermal treatment make it possible to prevent the formation of the zirconia layer, but they do not eliminate the hardening effect produced by oxygen diffusion. On the other hand, electrolytically produced chromium deposits whose adherence is improved by a thermal vacuum treatment, counteract both these phenomena simultaneously. A similar result has been obtained with coatings of molybdenum produced by the technique of high-frequency inductive plasma sputtering. The particular effectiveness of the last two types of coatings is due to their structures characterized by the existence of an adherent film of chromium or molybdenum in the free state. (authors) [fr

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

Science.gov (United States)

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

2018-02-01

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

Electrothermal atomic absorption spectrometric determination of copper in nickel-base alloys with various chemical modifiers*1

Science.gov (United States)

Tsai, Suh-Jen Jane; Shiue, Chia-Chann; Chang, Shiow-Ing

1997-07-01

The analytical characteristics of copper in nickel-base alloys have been investigated with electrothermal atomic absorption spectrometry. Deuterium background correction was employed. The effects of various chemical modifiers on the analysis of copper were investigated. Organic modifiers which included 2-(5-bromo-2-pyridylazo)-5-(diethylamino-phenol) (Br-PADAP), ammonium citrate, 1-(2-pyridylazo)-naphthol, 4-(2-pyridylazo)resorcinol, ethylenediaminetetraacetic acid and Triton X-100 were studied. Inorganic modifiers palladium nitrate, magnesium nitrate, aluminum chloride, ammonium dihydrogen phosphate, hydrogen peroxide and potassium nitrate were also applied in this work. In addition, zirconium hydroxide and ammonium hydroxide precipitation methods have also been studied. Interference effects were effectively reduced with Br-PADAP modifier. Aqueous standards were used to construct the calibration curves. The detection limit was 1.9 pg. Standard reference materials of nickel-base alloys were used to evaluate the accuracy of the proposed method. The copper contents determined with the proposed method agreed closely with the certified values of the reference materials. The recoveries were within the range 90-100% with relative standard deviation of less than 10%. Good precision was obtained.

The elastic properties of zirconium alloy fuel cladding and pressure tubing materials

International Nuclear Information System (INIS)

Rosinger, H.E.; Northwood, D.O.

1979-01-01

A knowledge of the elastic properties of zirconium alloys is required in the mathematical modelling of cladding and pressure tubing performance. Until recently, little of this type of data was available, particularly at elevated temperatures. The dynamic elastic moduli of zircaloy-2, zircaloy-4, the alloys Zr-1.0 wt%Nb, Zr-2.5 wt%Nb and Marz grade zirconium have therefore been determined over the temperature range 275 to 1000 K. Young's modulus and shear modulus for all the zirconium alloys decrease with temperature and are expressed by empirical relations fitted to the data. The elastic properties are texture dependent and a detailed study has been conducted on the effect of texture on the elastic properties of Zr-1.0 wt% Nb over the temperature range 275 to 775 K. The results are compared with polycrystalline elastic constants computed from single crystal elastic constants, and the effect of texture on the dynamic elastic moduli is discussed in detail. (Auth.)

New zirconium alloys for nuclear application; Novas ligas de zirconio para aplicacao nuclear

Energy Technology Data Exchange (ETDEWEB)

Lobo, R.M.; Andrade, A.H.P., E-mail: rmlobo@ipen.b [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

2010-07-01

Zirconium alloys are widely used in the nuclear industry, mainly in fuel cladding tubes and structural components for PWR plants. The service life of these components, which operate under high temperatures conditions ({approx} 300 deg C), has led to developing new alloys with the aim to improve the mechanical properties, corrosion resistance and irradiation damage. The variation in the composition of the alloy produces second phase particles which alter the materials properties according to their size and distribution, is essential therefore, knowledge their characteristics. Analysis of second phase particles in zirconium alloys are carried out by scanning electron microscopy, transmission electron microscopy and image analysis. This study used the zircaloy-4 to illustrate the characterization of these alloys through the study of second phase particles. (author)

Nickel and cobalt base alloys

International Nuclear Information System (INIS)

Houlle, P.

1994-01-01

Nickel base alloys have a good resistance to pitting, cavernous or cracks corrosion. Nevertheless, all the nickel base alloys are not equivalent. Some differences exit between all the families (Ni, Ni-Cu, Ni-Cr-Fe, Ni-Cr-Fe-Mo/W-Cu, Ni-Cr-Mo/W, Ni-Mo). Cobalt base alloys in corrosive conditions are generally used for its wear and cracks resistance, with a compromise to its localised corrosion resistance properties. The choice must be done from the perfect knowledge of the corrosive medium and of the alloys characteristics (chemical, metallurgical). A synthesis of the corrosion resistance in three medium (6% FeCl 3 , 4% NaCl + 1% HCl + 0.1% Fe 2 (SO 4 ) 3 , 11.5% H 2 SO 4 + 1.2% HCl + 1% Fe 2 (SO 4 ) 3 + 1% CuCl 2 ) is presented. (A.B.). 11 refs., 1 fig., 12 tabs

Nickel, cobalt, and their alloys

CERN Document Server

2000-01-01

This book is a comprehensive guide to the compositions, properties, processing, performance, and applications of nickel, cobalt, and their alloys. It includes all of the essential information contained in the ASM Handbook series, as well as new or updated coverage in many areas in the nickel, cobalt, and related industries.

High energy beam thermal processing of alpha zirconium alloys and the resulting articles

International Nuclear Information System (INIS)

Sabol, G.P.; McDonald, S.G.; Nurminen, J.I.

1983-01-01

Alpha zirconium alloy fabrication methods and resultant products exhibiting improved high temperature, high pressure steam corrosion resistance. The process, according to one aspect of this invention, utilizes a high energy beam thermal treatment to provide a layer of beta treated microstructure on an alpha zirconium alloy intermediate product. The treated product is then alpha worked to final size. According to another aspect of the invention, high energy beam thermal treatment is used to produce an alpha annealed microstructure in a Zircaloy alloy intermediate size or final size component. The resultant products are suitable for use in pressurized water and boiling water reactors

Corrosion properties of plasma deposited nickel and nickel-based alloys

Czech Academy of Sciences Publication Activity Database

Voleník, Karel; Pražák, M.; Kalabisová, E.; Kreislová, K.; Had, J.; Neufuss, Karel

2003-01-01

Roč. 48, č. 3 (2003), s. 215-226 ISSN 0001-7043 R&D Projects: GA ČR GA106/99/0298 Institutional research plan: CEZ:AV0Z2043910 Keywords : plasma deposits, nickel, nickel-based alloys Subject RIV: JK - Corrosion ; Surface Treatment of Materials

Effect of Ni +-ION bombardment on nickel and binary nickel alloys

Science.gov (United States)

Roarty, K. B.; Sprague, J. A.; Johnson, R. A.; Smidt, F. A.

1981-03-01

Pure nickel and four binary nickel alloys have been subjected to high energy Ni ion bombardment at 675, 625 and 525°C. After irradiation, each specimen was studied by transmission electron microscopy. The pure nickel control was found to swell appreciably (1 to 5%) and the Ni-Al and the Ni-Ti samples were found to swell at all temperatures, but to a lesser degree (0.01 to 0.35%). The Ni-Mo contained a significant density of voids only at 525° C, while swelling was suppressed at all temperatures in the Ni-Si alloy. The dislocation structure progressed from loops to tangles as temperature increased in all materials except the Ni-Ti, in which there was an absence of loops at all temperatures. Dislocation densities decreased as temperature increased in all samples. These results do not correlate well with the relative behavior of the same alloys observed after neutron irradiation at 455°C. The differences between these two sets of data appear to be caused by different mechanisms controlling void nucleation in ion and neutron irradiation of these alloys.

Environmentally-induced cracking of zirconium alloys - a review

International Nuclear Information System (INIS)

Cox, B.

1990-01-01

The general field of environmentally-induced cracking of zirconium alloys has been reviewed and the phenomena that are observed and the progress in understanding the mechanisms are summarized. The details of the industrially important pellet-clad interaction failures of nuclear reactor fuel have been left for a companion review, and only observations on the mechanism are summarized briefly here. It is concluded that in the zirconium alloy system, by virtue of the physical peculiarities of the system, it is easier to reach unambiguous conclusions about the environmental cracking mechanisms that are operating than with other systems. Thus, chemical dissolution in either liquid or vapour phase is thought to be the principal mechanism for intergranular cracking, while adsorption-induced embrittlement is thought to be the most common transgranular quasi-cleavage process. Hydrogen embrittlement in this system can be identified because it requires precipitated hydride that gives characteristic fractography when cracked. Only in a few instances does stress-corrosion cracking appear to proceed by a hydride cracking mechanism. (orig.)

Copper and copper-nickel-alloys - An overview

Energy Technology Data Exchange (ETDEWEB)

Klassert, Anton; Tikana, Ladji [Deutsches Kupferinstitut e.V. Am Bonneshof 5, 40474 Duesseldorf (Germany)

2004-07-01

With the increasing level of industrialization the demand for and the number of copper alloys rose in an uninterrupted way. Today, the copper alloys take an important position amongst metallic materials due to the large variety of their technological properties and applications. Nowadays there exist over 3.000 standardized alloys. Copper takes the third place of all metals with a worldwide consumption of over 15 millions tons per year, following only to steel and aluminum. In a modern industrial society we meet copper in all ranges of the life (electro-technology, building and construction industry, mechanical engineering, automotive, chemistry, offshore, marine engineering, medical applications and others.). Copper is the first metal customized by humanity. Its name is attributed to the island Cyprus, which supplied in the antiquity copper to Greece, Rome and the other Mediterranean countries. The Romans called it 'ore from Cyprus' (aes cyprium), later cuprum. Copper deposited occasionally also dapper and could be processed in the recent stone age simply by hammering. Already in early historical time copper alloys with 20 to 50 percent tin was used for the production of mirrors because of their high reflecting power. Although the elementary nickel is an element discovered only recently from a historical perspective, its application in alloys - without any knowledge of the alloy composition - occurred at least throughout the last 2.000 years. The oldest copper-nickel coin originates from the time around 235 B.C.. Only around 1800 AD nickel was isolated as a metallic element. In particular in the sea and offshore technology copper nickel alloys found a broad field of applications in piping systems and for valves and armatures. The excellent combination of characteristics like corrosion resistance, erosion stability and bio-fouling resistance with excellent mechanical strength are at the basis of this success. An experience of many decades supports the use

The GENIALL process for generation of nickel-iron alloys from nickel ores or mattes

International Nuclear Information System (INIS)

Diaz, G.; Frias, C.; Palma, J.

2001-01-01

A new process, called GENIALL (acronym of Generation of Nickel Alloys), for nickel recovery as ferronickel alloys from ores or mattes without previous smelting is presented in this paper. Its core technology is a new electrolytic concept, the ROSEL cell, for electrowinning of nickel-iron alloys from concentrated chloride solutions. In the GENIALL Process the substitution of iron-based solid wastes as jarosite, goethite or hematite, by saleable ferronickel plates provides both economic and environmental attractiveness. Another advantage is that no associated sulfuric acid plant is required. The process starts with leaching of the raw material (ores or mattes) with a solution of ferric chloride. The leachate liquor is purified by conventional methods like cementation or solvent extraction, to remove impurities or separate by-products like copper and cobalt. The purified solution, that contains a mixture of ferrous and nickel chlorides is fed to the cathodic compartment of the electrowinning cell, where nickel and ferrous ions are reduced together to form an alloy. Simultaneously, ferrous chloride is oxidized to ferric chloride in the anodic compartment, from where it is recycled to the leaching stage. The new electrolytic equipment has been developed and scaled up from laboratory to pilot prototypes with commercial size electrodes of 1 m 2 . Process operating conditions have been established in continuous runs at bench and pilot plant scale. The technology has shown a remarkable capacity to produce nickel-iron alloys of a wide range of compositions, from 10% to 80% nickel, just by adjusting the operating parameters. This emerging technology could be implemented in many processes in which iron and other non-ferrous metals are harmful impurities to be removed, or valuable metals to be recovered as a marketable iron alloy. Other potential applications of this technology are regeneration of spent etching liquors, and iron removal from aqueous effluents. (author)

Corrosion of zirconium alloys in alternating pH environment

International Nuclear Information System (INIS)

Mayer, P.; Manolescu, A.V.

1985-01-01

Behaviour of two commercial alloys, Zircaloy-2 and zirconium-2.5 wt% niobium were investigated in an environment of alternating pH. Corrosion advancement and scale morphology of coupons exposed to aqueous solution of LiOH (pH 10.2 and 14) were followed as a function of temperature (300-360 degreesC) and time (up to 165 days). The test sequence consisted of short term exposure to high pH and re-exposure to low pH solutions for extended period of time followed by a short term test in high pH. The results of these tests and detailed post-corrosion analysis indicate a fundamental difference between the corrosion behaviour of these two materials. Both alloys corrode fast in high pH environments, but only zirconium-2.5 wt% niobium continues to form detectable new oxide in low pH solution

Zinc-nickel alloy electrodeposits for water electrolysis

Energy Technology Data Exchange (ETDEWEB)

Sheela, G.; Pushpavanam, Malathy; Pushpavanam, S. [Central Electrochemical Research Inst., Karaikudi (India)

2002-06-01

Electrodeposited zinc-nickel alloys of various compositions were prepared. A suitable electrolyte and conditions to produce alloys of various compositions were identified. Alloys produced on electroformed nickel foils were etched in caustic to leach out zinc and to produce the Raney type, porous electro catalytic surface for hydrogen evolution. The electrodes were examined by polarisation measurements, to evaluate their Tafel parameters, cyclic voltammetry, to test the change in surface properties on repeated cycling, scanning electron microscopy to identify their microstructure and X-ray diffraction. The catalytic activity as well as the life of the electrode produced from 50% zinc alloy was found to be better than others. (Author)

Proceedings [of the] symposium on zirconium alloys for reactor components

International Nuclear Information System (INIS)

1992-01-01

A two day symposium on zirconium alloys for reactor components (ZARC-91) was organised during 12-13, 1991. There were 6 invited talks and 43 contributed papers in 10 technical sessions. This symposium, took stock of the progress achieved in the development, design, fabrication and quality assurance of zirconium alloy components and emphasized the R and D efforts required for meeting the challenges posed by the rapid growth of nuclear power in our country. Topics like physical metallurgy, corrosion and irradiation behaviour, and in-service inspection were also covered. The proceedings/papers are arranged under the headings: (1)invited talks, (2)fabrication, (3)design requirement, (4)quality assurance, (5)irradiation damage and PIE, (6)corrosion and hydriding, and (7)in-service inspection. (N.B.). refs., figs., tabs

Void formation in irradiated binary nickel alloys

International Nuclear Information System (INIS)

Shaikh, M.A.; Ahmed, M.; Akhter, J.I.

1994-01-01

In this work a computer program has been used to compute void radius, void density and swelling parameter for nickel and binary nickel-carbon alloys irradiated with nickel ions of 100 keV. The aim is to compare the computed results with experimental results already reported

Alloys of nickel-iron and nickel-silicon do not swell under fast neutron irradiation

International Nuclear Information System (INIS)

Silvestre, G.; Silvent, A.; Regnard, C.; Sainfort, G.

1975-01-01

This research is concerned with the effect of fast-neutron irradiation on the swelling of nickel and nickel alloys. Ni-Fe (0-60at%Fe) and Ni-Si (0-8at%Si) were studied, and the fluences were in the range 10 20 -4.3x10 22 n/cm 2 . In dilute alloys, the added elements are dissolved and reduce swelling, silicon being particularly effective. In more concentrated alloys, irradiation of Ni-Fe and Ni-Si alloys brings about the formation of plate-shaped precipitates of Ni 3 X and these alloys do not swell. (Auth.)

Development of Self-Healing Zirconium-Silicide Coatings for Improved Performance Zirconium-Alloy Fuel Cladding

Energy Technology Data Exchange (ETDEWEB)

Sridharan, Kumar [University of Wisconsin-Madison; Mariani, Robert [Idaho National Lab. (INL), Idaho Falls, ID (United States); Bai, Xianming [Idaho National Lab. (INL), Idaho Falls, ID (United States); Xu, Peng [Westinghouse Electric Company; Lahoda, Ed [Westinghouse Electric Company

2018-03-31

Given the long-term goal of developing such coatings for use with nuclear reactor fuel cladding, this work describes results of oxidation and corrosion behavior of bulk zirconium-silicide and fabrication of zirconium-silicide coatings on zirconium-alloy test flats, tube configurations, and SiC test flats. In addition, boiling heat transfer of these modified surfaces (including ZrSi2 coating) during clad quenching experiments is discussed in detail. Oxidation of bulk ZrSi2 was found to be negligible compared to Zircaloy-4 (a common Zr-alloy cladding material) and mechanical integrity of ZrSi2 was superior to that of bulk Zr2Si at high temperatures in ambient air. Very interesting and unique multi-nanolayered composite of ZrO2 and SiO2 were observed. Physical model for the oxidation has been proposed wherein Zr–Si–O mixture undergoes a spinodal phase decomposition into ZrO2 and SiO2, which is manifested as a nanoscale assembly of alternating layer of the two oxides. Steam corrosion at high pressure (10.3 MPa) led to weight loss of ZrSi2 and produced oxide scale with depletion of silicon, possibly attributed to volatile silicon hydroxide, gaseous silicon monoxide, and a solubility of silicon dioxide in water. Only Zircon phase (ZrSiO4) formed during oxidation of ZrSi2 at 1400°C in air, and allowed for immobilization silicon species in oxide scale in the aqueous environments. Zirconium-silicide coatings (on zirconium-alloy substrates) investigated in this study were deposited primarily using magnetron sputter deposition method and slurry method, although powder spray deposition processes cold spray and thermal spray methods were also investigated. The optimized ZrSi2 sputtered coating exhibited a highly protective nature at elevated temperatures in ambient air by mitigating oxygen permeation to the underlying zirconium alloy substrate. The high oxidation resistance of the coating has been shown to be due to nanocrystalline SiO2 and ZrSiO4 phases in the amorphous

Traps in Zirconium Alloys Oxide Layers

Directory of Open Access Journals (Sweden)

Helmar Frank

2005-01-01

Full Text Available Oxide films long-time grown on tubes of three types of zirconium alloys in water and in steam were investigated, by analysing I-V characteristic measured at constant voltages with various temperatures. Using theoretical concepts of Rose [3] and Gould [5], ZryNbSn(Fe proved to have an exponential distribution of trapping centers below the conduction band edge, wheras Zr1Nb and IMP Zry-4 proved to have single energy trap levels.

The oxidation kinetics of zirconium alloys applicable to loss-of-coolant accidents

International Nuclear Information System (INIS)

Parsons, P.D.; Miller, W.N.

1977-10-01

A review is presented of the available published measurements of the rate of reaction between zirconium alloys and steam and, in some cases, oxygen. Attempts are made to define from all the experimental data a suitable rate equation which is appropriate over the range of temperatures relevant to LOCA conditions. The data reviewed encompass a temperature range 910 0 C to the melting point of zirconium, 1852 0 C. It can be concluded that within 910 to 1577 0 C, Zircaloy-2, Zircaloy-4 and Zr/2 1/2%Nb alloys have the same response to oxidation. (author)

High-intensity low energy titanium ion implantation into zirconium alloy

Science.gov (United States)

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.

Ductile tungsten-nickel alloy and method for making same

Science.gov (United States)

Snyder, Jr., William B.

1976-01-01

The present invention is directed to a ductile, high-density tungsten-nickel alloy which possesses a tensile strength in the range of 100,000 to 140,000 psi and a tensile elongation of 3.1 to 16.5 percent in 1 inch at 25.degree.C. This alloy is prepared by the steps of liquid phase sintering a mixture of tungsten-0.5 to 10.0 weight percent nickel, heat treating the alloy at a temperature above the ordering temperature of approximately 970.degree.C. to stabilize the matrix phase, and thereafter rapidly quenching the alloy in a suitable liquid to maintain the matrix phase in a metastable, face-centered cubic, solid- solution of tungsten in nickel.

Plate-shaped transformation products in zirconium-base alloys

International Nuclear Information System (INIS)

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

1997-01-01

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

Study of diffusion processes in the oxide layer of zirconium alloys

Directory of Open Access Journals (Sweden)

Sialini P.

2016-03-01

Full Text Available In the active zone of a nuclear reactor where zirconium alloys are used as a coating material, this material is subject to various harmful impacts. During water decomposition reactions, hydrogen and oxygen are evolved that may diffuse through the oxidic layer either through zirconium dioxide (ZrO2 crystals or along ZrO2 grains. The diffusion mechanism can be studied using the Ion Beam Analysis (IBA method where nuclear reaction 18O(p,α15N is used. A tube made of zirconium alloy E110 (with 1 wt. % of Nb was used for making samples that were pre-exposed in UJP PRAHA a.s. and subsequently exposed to isotopically cleansed environment of H2 18O medium in an autoclave. The samples were analysed with gravimetric methods and IBA methods performed at the electrostatic particle accelerator Tandetron 4130 MC in the Nucler Physics Institute of the CAS, Řež. With IBA methods, the overall thicknesses of corrosion layers on the samples, element composition of the alloy and distribution of oxygen isotope 18O in the corrosion layer and its penetration in the alloy were identified. The retrieved data shows at the oxygen diffusion along ZrO2 grains because there are two peaks of 18O isotope concentrations in the corrosion layer. These peaks occur at the environment-oxide and oxide-metal interface. The element analysis identified the presence of undesirable hafnium.

A sulfidation-resistant nickel-base alloy

International Nuclear Information System (INIS)

Lai, G.Y.

1989-01-01

For applications in mildly to moderately sulfidizing environments, stainless steels, Fe-Ni-Cr alloys (e.g., alloys 800 and 330), and more recently Fe-Ni-Cr-Co alloys (e.g., alloy 556) are frequently used for construction of process equipment. However, for many highly sulfidizing environments, few existing commercial alloys have adequate performance. Thus, a new nickel-based alloy containing 27 wt.% Co, 28 wt.% Cr, 4 wt.% Fe, 2.75 wt.% Si, 0.5 wt.% Mn and 0.05 wt.% C (Haynes alloy HR-160) was developed

A microstuctural study on accelerated zirconium alloy oxidation

International Nuclear Information System (INIS)

Sohn, Seung Bum; Oh, Seung Jun; Jang, Jung Nam; Kim, Yong Soo; Jung, Yong Hwan; Baek, Jong Hyuk; Park, Jung Yong

2005-01-01

It has been reported that the effect of thermal redistribution of hydrides across the zirconium metaloxide interface, coupled with thermal feedback on the metal-oxide interface, is a dominating factor in the accelerated oxidation in zirconium alloys cladding PWR fuel. Basically this influence determines characteristic of oxide layer. Influence estimation for corrosion oxide layer due to hydrogen / hydride carried out because of investigation on the kinetic on accelerated oxidation due to hydride precipitation was preceded. Generally, it is known that ZrO 2 tetragonal layer structures play an important role as a barrier layer. So analysing the ZrO 2 monoclinic and tetragonal structure distribution is our main aim. Especially, this study focused on the hydride effects. In other words, the difference of crystal structure distribution between pre-hydrided and without hydrided specimen is just expected results. Experimental results of microstructure at zirconium metal-oxide interface through TEM and EBSD analysis was confirmed

Methods for determination of zirconium in titanium alloys

International Nuclear Information System (INIS)

1985-01-01

Two methods for determining zirconium content in titanium alloys are specified in this standard. One is the ion-exchange/mandelic acid gravimetry for Zr content below 20 % down to 1 % while the other is the mandelic acid gravimetry for Zr content below 20 % down to 0.5 %. In the former, a specimen is decomposed by hydrochloric acid and hydrofluoric acid. After substances such as titanium are oxidized by adding nitric acid, the liquid is adjusted into a 4N hydrochloric acid - gN hydrofluoric acid solution, which is them passed through an ion-exchange column. The niobium and tantalum contents are absorbed while the titanium and zirconium contents flow out. Perchloric acid and sulfuric acid are poured in the solution to remove hydrofluoric acid. Aqueous ammonia is added to produce hydroxide of titanium and zirconium, which is then filtered out. The hydroxyde is dissolved in hydrochloric acid, and mandelic acid is poured to precipitate the zirconium content. The precipitate is ignited and the weight of the oxide formed is measured. The coprecipitated titanium content is determined by the absorptiometric method using hydrogen peroxide. Finally, the weight of the oxide is corrected. In the latter determination method, on the other hand, only several steps of the above procedure are used, namely, decomposition by hydrochloric acid, precipitation of zirconium, ignition of precipitate, measurement of oxide weight and weight correction. (Nogami, K.)

Analysis of weld solidification cracking in cast nickel aluminide alloys

International Nuclear Information System (INIS)

Santella, M.L.; Feng, Z.

1995-01-01

A study of the response of several nickel aluminide alloys to SigmaJig testing was done to examine their weld solidification cracking behavior and the effect of Zr concentration. The alloys were based on the Ni-8Al-7.7Cr-1.5Mo-0.003B wt% composition and contained Zr concentrations of 3, 4.5, and 6 wt%. Vacuum induction melted ingots with a diameter of 2.7 in and weight about 18 lb were made of each alloy, and were used to make 2 x 2 x 0.030 in specimens for the Sigmajig test. The gas tungsten arc welds were made at travel speeds of 10, 20, and 30 ipm with heat inputs of 2--2.5 kJ/in. When an arc was established before traveling onto the test specimen centerline cracking was always observed. This problem was overcome by initiating the arc directly on the specimens. Using this approach, the 3 wt% Zr alloy withstood an applied stress of 24 ksi without cracking at a welding speed of 10 ipm. This alloy cracked at 4 ksi applied at 20 ipm, and with no applied load at 30 ipm. Only limited testing was done on the remaining alloys, but the results indicate that resistance to solidification cracking increases with Zr concentration. Zirconium has limited solid solubility and segregates strongly to interdendritic regions during solidification where it forms a Ni solid solution-Ni 5 Zr eutectic. The volume fraction of the eutectic increases with Zr concentration. The solidification cracking behavior of these alloys is consistent with phenomenological theory, and is discussed in this context. The results from SigmaJig testing are analyzed using finite element modeling of the development of mechanical strains during solidification of welds. Experimental data from the test substantially agree with recent analysis results

Nickel aluminide alloy suitable for structural applications

Science.gov (United States)

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.

Standard specification for Nickel-Chromium-Iron alloys (UNS N06600, N06601, N06603, N06690, N06693, N06025, N06045 and N06696), Nickel-Chromium-Cobalt-Molybdenum alloy (UNS N06617), and Nickel-Iron-Chromium-Tungsten alloy (UNS N06674) plate, sheet and strip

CERN Document Server

American Society for Testing and Materials. Philadelphia

2011-01-01

Standard specification for Nickel-Chromium-Iron alloys (UNS N06600, N06601, N06603, N06690, N06693, N06025, N06045 and N06696), Nickel-Chromium-Cobalt-Molybdenum alloy (UNS N06617), and Nickel-Iron-Chromium-Tungsten alloy (UNS N06674) plate, sheet and strip

Standard specification for Nickel-Chromium-Iron alloys (UNS N06600, N06601, N06603, N06690, N06693, N06025, N06045, and N06696), Nickel-Chromium-Cobalt-Molybdenum alloy (UNS N06617), and Nickel-Iron-Chromium-Tungsten alloy (UNS N06674) rod, bar, and wire

CERN Document Server

American Society for Testing and Materials. Philadelphia

2011-01-01

Standard specification for Nickel-Chromium-Iron alloys (UNS N06600, N06601, N06603, N06690, N06693, N06025, N06045, and N06696), Nickel-Chromium-Cobalt-Molybdenum alloy (UNS N06617), and Nickel-Iron-Chromium-Tungsten alloy (UNS N06674) rod, bar, and wire

Electroless nickel-plating for the PWSCC mitigation of nickel-base alloys in nuclear power plants

International Nuclear Information System (INIS)

Kim, Ji Hyun; Hwang, Il Soon

2008-01-01

The feasibility study has been performed as an effort to apply the electroless nickel-plating method for a proposed countermeasure to mitigate primary water stress corrosion cracking (PWSCC) of nickel-base alloys in nuclear power plants. In order to understand the corrosion behavior of nickel-plating at high temperature water, the electrochemical properties of electroless nickel-plated alloy 600 specimens exposed to simulated pressurized water reactor (PWR) primary water were experimentally characterized in high temperature and high pressure water condition. And, the resistance to the flow accelerated corrosion (FAC) test was investigated to check the durability of plated layers in high-velocity water-flowing environment at high temperature. The plated surfaces were examined by using both scanning electron microscopy (SEM) and energy dispersive spectroscopy (EDS) after exposures to the condition. From this study, it is found that the corrosion resistance of electroless nickel-plated Alloy 600 is higher than that of electrolytic plating in 290 deg. C water

Carbon formation on nickel and nickel-copper alloy catalysts

Energy Technology Data Exchange (ETDEWEB)

Alstrup, I.; Soerensen, O.; Rostrup-Nielsen, J.R. [Haldor Topsoe Research Labs., Lyngby (Denmark); Tavares, M.T.; Bernardo, C.A.

1998-05-01

Equilibrium, kinetic and morphological studies of carbon formation in CH{sub 4} + H{sub 2}, CO, and CO + H{sub 2} gases on silica supported nickel and nickel-copper catalysts are reviewed. The equilibrium deviates in all cases from graphite equilibrium and more so in CO + CO{sub 2} than in CH{sub 4} + H{sub 2}. A kinetic model based on information from surface science results with chemisorption of CH{sub 4} and possibly also the first dehydrogenation step as rate controlling describes carbon formation on nickel catalyst in CH{sub 4} + H{sub 2} well. The kinetics of carbon formation in CO and CO + H{sub 2} gases are in agreement with CO disproportionation as rate determining step. The presence of hydrogen influences strongly the chemisorption of CO. Carbon filaments are formed when hydrogen is present in the gas while encapsulating carbon dominates in pure CO. Small amounts of Cu alloying promotes while larger amounts (Cu : Ni {>=} 0.1) inhibits carbon formation and changes the morphology of the filaments (``octopus`` carbon formation). Adsorption induced nickel segregation changes the kinetics of the alloy catalysts at high carbon activities. Modifications suggested in some very recent papers on the basis of new results are also briefly discussed. (orig.) 31 refs.

The Effect of Hydrogen and Hydrides on the Integrity of Zirconium Alloy Components Delayed Hydride Cracking

CERN Document Server

Puls, Manfred P

2012-01-01

By drawing together the current theoretical and experimental understanding of the phenomena of delayed hydride cracking (DHC) in zirconium alloys, The Effect of Hydrogen and Hydrides on the Integrity of Zirconium Alloy Components: Delayed Hydride Cracking provides a detailed explanation focusing on the properties of hydrogen and hydrides in these alloys. Whilst the focus lies on zirconium alloys, the combination of both the empirical and mechanistic approaches creates a solid understanding that can also be applied to other hydride forming metals.   This up-to-date reference focuses on documented research surrounding DHC, including current methodologies for design and assessment of the results of periodic in-service inspections of pressure tubes in nuclear reactors. Emphasis is placed on showing that our understanding of DHC is supported by progress across a broad range of fields. These include hysteresis associated with first-order phase transformations; phase relationships in coherent crystalline metallic...

Diffusion model of delayed hydride cracking in zirconium alloys

NARCIS (Netherlands)

Shmakov, AA; Kalin, BA; Matvienko, YG; Singh, RN; De, PK

2004-01-01

We develop a method for the evaluation of the rate of delayed hydride cracking in zirconium alloys. The model is based on the stationary solution of the phenomenological diffusion equation and the detailed analysis of the distribution of hydrostatic stresses in the plane of a sharp tensile crack.

Effects of ion implantation on corrosion of zirconium and zirconium base alloys

International Nuclear Information System (INIS)

Zelenskij, V.F.; Petel'guzov, I.A.; Rekova, L.P.; Rodak, A.G.

1989-01-01

The influence of He and Ar ion bombardment on the corrosion of Zr and Zr-1%Nb and Zr-2.5%Nb alloys is investigated with the aims of finding the irradiation influence laws, obtaining the dependences of the effect of increasing the corrosiuon resistance on the type and dose of bombarding ions and of finding the conditions for the maximum effect. The prolonged corrosion test of specimens (3500 hours) have shown that the strongest effect is obtained for the irradiation with Ar ions up to the dose 1x10 16 ion/cm 2 . The kinetics of ion thermosorption after corrosion of irradiated materials is studied, the temperature threshold of implanted ion stability in zirconium and its alloys is found to be 400 deg C

Copper and nickel alloys and titanium for seawater applications

International Nuclear Information System (INIS)

Richter, H.

1977-01-01

Copper and nickel alloys and titanium have been successfully used for heat exchangers on ships, in power plants and for chemical apparatus and piping systems because of their resistance against corrosion in sea water. Aluminium brass and copper nickel alloys, the standard materials for condensers and coolers, however, may be attacked, the corrosion depending on water quality, water velocity, and structural conditions. The mechanisms of corrosion are discussed. Under severe conditions the use of titanium may be indicated. The use of nickel base alloys is advantageous at elevated temperatures, e.g. for chemical reactions and for evaporation processes. Examples are given for application and for prevention of corrosion. (orig.) [de

Mechanistic understanding of irradiation corrosion of zirconium alloys in nuclear power plants: stimuli, status and outlook

International Nuclear Information System (INIS)

Cox, B.; Ishigure, K.; Johnson, A.B.; Lemalgnan, J.C.; Nechaev, A.F.; Petrik, N.G.; Reznichenko, E.A.

1990-01-01

Extensive information about the corrosion behaviour of zirconium alloys under irradiation is presented. Review of the existing models of radiation corrosion is given. An accent is made on a necessity in conducting basic investigations to overcome contradictions in interpreting the experimental data available. Importance of solving the problem of zirconium alloy corrosion for safe NPP operation is underlined. 34 refs.; 6 figs.; 4 tabs

Oxidation resistance of nickel alloys at high temperature

International Nuclear Information System (INIS)

Tyuvin, Yu.D.; Rogel'berg, I.L.; Ryabkina, M.M.; Plakushchaya, A.F.

1977-01-01

The heat resistance properties of nickel alloys Ni-Cr-Si, Ni-Si-Al, Ni-Si-Mn and Ni-Al-Mn have been studied by the weight method during oxidation in air at 1000 deg and 1200 deg C. It is demonstrated that manganese reduces the heat resistance properties of Ni-Si and Ni-Al alloys, whilst the addition of over 3% aluminium enhances the heat resistance properties of Ni-Si (over 1.5%) alloys. The maximum heat resistance properties are shown by Ni-Si-Al and Ni-Cr-Si alloys with over 2% Si. These alloys offer 3 to 4 times better oxidation resistance as compared with pure nickel at 1000 deg C and 10 times at 1200 deg C

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

Energy Technology Data Exchange (ETDEWEB)

Sindelar, R. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Louthan, M. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); PNNL, B. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

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, 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 to

Strengthening and elongation mechanism of Lanthanum-doped Titanium-Zirconium-Molybdenum alloy

Energy Technology Data Exchange (ETDEWEB)

Hu, Ping, E-mail: huping1985@126.com [School of Metallurgy Engineering, Xi’an University of Architecture and Technology, Xi’an 710055 (China); Jinduicheng Molybdenum Co., Ltd., Xi’an 710068 (China); Hu, Bo-liang; Wang, Kuai-she; Song, Rui; Yang, Fan [School of Metallurgy Engineering, Xi’an University of Architecture and Technology, Xi’an 710055 (China); Yu, Zhi-tao [Ruifulai Tungsten & Molybdenum Co., Ltd., Xi’an 721914 (China); Tan, Jiang-fei [School of Metallurgy Engineering, Xi’an University of Architecture and Technology, Xi’an 710055 (China); Cao, Wei-cheng; Liu, Dong-xin; An, Geng [Jinduicheng Molybdenum Co., Ltd., Xi’an 710068 (China); Guo, Lei [Ruifulai Tungsten & Molybdenum Co., Ltd., Xi’an 721914 (China); Yu, Hai-liang [School of Mechanical, Materials and Mechatronics Engineering, University of Wollongong, NSW 2522 (Australia)

2016-12-15

The microstructural contributes to understand the strengthening and elongation mechanism in Lanthanum-doped Titanium-Zirconium-Molybdenum alloy. Lanthanum oxide particles not only act as heterogeneous nucleation core, but also act as the second phase to hinder the grain growth during sintering crystallization. The molybdenum substrate formed sub-grain under the effect of second phase when the alloy rolled to plate.

Optimization of the composition and structure of heat-resistant casting aluminium alloys with additions of cerium, iron, nickel and zirconium

International Nuclear Information System (INIS)

Belov, N.A.; Lavrishchev, Yu.V.

2000-01-01

A study is made of the effect of composition and structure on mechanical properties of cast alloys of the Al-Ce-Ni-Fe-Zr system in which binary and ternary eutectics with participation of low alloyed aluminium solid solution and Al 4 Ce, Al 3 Ni and Al 9 FeNi phases are crystallized. It is found that microhardness of eutectics is heavily dependent on the volume fraction of aluminides and their dispersivity. It was shown that essential hardening of aluminium matrix can be achieved at the cost of zirconium additive in quantity of 0.6 % when using two-stage manufacturing operation. Experimental compositions of Al-10 % Ce-5% Ni-0.6 % Zr and Al-1.5 % Fe-1.5 % Ni-0.6 % Zr on the basis of ternary and binary eutectics respectively as billets essentially exceed industrial heat-resistant cast aluminium alloys AK12MMgN and AM5 as to a set of room and high-temperature mechanical properties and hot brittleness index [ru

Minimizing hydride cracking in zirconium alloys

International Nuclear Information System (INIS)

Coleman, C.E.; Cheadle, B.A.; Ambler, J.F.R.; Eadie, R.L.

1985-01-01

Zirconium alloy components can fail by hydride cracking if they contain large flaws and are highly stressed. If cracking in such components is suspected, crack growth can be minimized by following two simple operating rules: components should be heated up from at least 30K below any operating temperature above 450K, and when the component requires cooling to room temperature from a high temperature, any tensile stress should be reduced as much and as quickly as is practical during cooling. This paper describes the physical basis for these rules

Nitrogen annealing of zirconium or titanium metals and their alloys

International Nuclear Information System (INIS)

Eucken, C.M.

1982-01-01

A method is described of continuously nitrogen annealing zirconium and titanium metals and their alloys at temperatures at from 525 0 to 875 0 C for from 1/2 minute to 15 minutes. The examples include the annealing of Zircaloy-4. (U.K.)

Effects of scandium and zirconium combination alloying on as-cast microstructure and mechanical properties of Al-4Cu-1.5Mg alloy

Directory of Open Access Journals (Sweden)

Xiang Qingchun

2011-02-01

Full Text Available The influences of minor scandium and zirconium combination alloying on the as-cast microstructure and mechanical properties of Al-4Cu-1.5Mg alloy have been experimentally investigated. The experimental results show that when the minor elements of scandium and zirconium are simultaneously added into the Al-4Cu-1.5Mg alloy, the as-cast microstructure of the alloy is effectively modified and the grains of the alloy are greatly refined. The coarse dendrites in the microstructure of the alloy without Sc and Zr additions are refined to the uniform and fine equiaxed grains. As the additions of Sc and Zr are 0.4% and 0.2%, respectively, the tensile strength, yield strength and elongation of the alloy are relatively better, which are 275.0 MPa, 176.0 MPa and 8.0% respectively. The tensile strength is increased by 55.3%, and the elongation is nearly raised three times, compared with those of the alloy without Sc and Zr additions.

Zirconium - an imported mineral commodity

International Nuclear Information System (INIS)

1983-10-01

This report examines Canada's position in regard to the principal zirconium materials: zircon; fusion-cast zirconium-bearing refractory products; zirconium-bearing chemicals; and zirconium metal, master alloys, and alloys. None of these is produced in Canada except fused alumina-zirconia and certain magnesium-zirconium alloys and zirconium-bearing steels. Most of the 3 000-4 000 tonnes of the various forms of zircon believed to be consumed in Canada each year is for foundry applications. Other minerals, notably chromite, olivine and silica sand are also used for these purposes and, if necessary, could be substituted for zircon. Zirconium's key role in Canada is in CANDU nuclear power reactors, where zirconium alloys are essential in the cladding for fuel bundles and in capital equipment such as pressure tubes, calandria tubes and reactivity control mechanisms. If zirconium alloys were to become unavailable, the Canadian nuclear power industry would collapse. As a contingency measure, Ontario Hydro maintains at least nine months' stocks of nuclear fuel bundles. Canada's vulnerability to short-term disruptions to supplies of nuclear fuel is diminished further by the availability of more expensive electricity from non-nuclear sources and, given time, from mothballed thermal plants. Zirconium minerals are present in many countries, notably Australia, the Republic of South Africa and the United States. Australia is Canada's principal source of zircon imports; South Africa is its sole source of baddeleyite. At this time, there are no shortages of either material. Canada has untapped zirconium resources in the Athabasca Oil Sands (zircon) and at Strange Lake along the ill-defined border between Quebec and Newfoundland (gittinsite). Adequate metal and alloy production facilities exist in France, Japan and the United States. No action by the federal government in regard to zirconium supplies is called for at this time

Waterside corrosion of zirconium alloys in nuclear power plants

International Nuclear Information System (INIS)

Jeong, Yong Hwan; Baek, B. J.; Park, S. Y. and others

1999-08-01

The overview of corrosion and hydriding behaviors of Zr-based alloy under the conditions of the in-reactor service and in the absence of irradiation is introduced in this report. The metallurgical characteristics of Zr-based alloys and the thermo-mechanical treatments on the microstructures and the textures in the manufacturing process for fuel cladding are also introduced. The factors affecting the corrosion of Zr alloy in reactor are summarized. And the corrosion mechanism and hydrogen up-take are discussed based on the laboratory and in-reactor results. The phenomenological observations of zirconium alloy corrosion in reactors are summarized and the models of in-reactor corrosion are exclusively discussed. Finally, the effects of irradiation on the corrosion process in Zr alloy were investigated mainly based on the literature data. (author). 538 refs., 26 tabs., 105 figs

Nickel-base alloys having a low coefficient of thermal expansion

International Nuclear Information System (INIS)

Baldwin, J.F.; Maxwell, D.H.

1975-01-01

Alloy compositions consisting predominantly of nickel, chromium, molybdenum, carbon, and boron are disclosed. The alloys possess a duplex structure consisting of a nickel--chromium--molybdenum matrix and a semi-continuous network of refractory carbides and borides. A combination of desirable properties is provided by these alloys, including elevated temperature strength, resistance to oxidation and hot corrosion, and a very low coefficient of thermal expansion

An overview of microstructural and experimental factors that affect the irradiation growth behavior of zirconium alloys

International Nuclear Information System (INIS)

Fidleris, V.; Tucker, R.P.; Adamson, R.B.

1987-01-01

This paper presents an overview of factors affecting irradiation growth of zirconium alloys. Recent data obtained from irradiation programs in EBR-II, ATR, and NRU reactors are used to illustrate the effects of various microstructural and experimental factors on the growth of Zircaloy, zirconium, and zirconium-biobium alloys irradiated to fluences up to 2 X 10 26 nm -2 (E > 1 MeV) over the temperature range 330 to 720 K. Open literature results are also used to confirm or illustrate various effects. Important factors are texture, grain boundary parameters, residual stresses, original dislocation density, microstructure evolution, temperature during irradiation, solute effects, and fluence

Austenitic stainless steel alloys with high nickel contents in high temperature liquid metal systems

International Nuclear Information System (INIS)

Konvicka, H.R.; Schwarz, N.F.

1981-01-01

Fe-Cr-Ni base alloys (nickel content: from 15 to 70 wt%, Chromium content: 15 wt%, iron: balance) together with stainless steel (W.Nr. 1.4981) have been exposed to flowing liquid sodium at 730 0 C in four intervals up to a cumulative exposure time of 1500 hours. Weight change data and the results of post-exposition microcharacterization of specimens are reported. The corrosion rates increase with increasing nickel content and tend to become constant after longer exposure times for each alloy. The corrosion rate of stainless steel is considerably reduced due to the presence of the base alloys. Different kinetics of nickel poor (up to 35% nickel) and nickel rich (> 50% nickel) alloys and nickel transport from nickel rich to nickel poor material is observed. (orig.)

Nickel-titanium alloys: a systematic review

Directory of Open Access Journals (Sweden)

Marcelo do Amaral Ferreira

2012-06-01

Full Text Available OBJECTIVE: A systematic review on nickel-titanium wires was performed. The strategy was focused on Entrez-PubMed-OLDMEDLINE, Scopus and BioMed Central from 1963 to 2008. METHODS: Papers in English and French describing the behavior of these wires and laboratorial methods to identify crystalline transformation were considered. A total of 29 papers were selected. RESULTS: Nickel-titanium wires show exceptional features in terms of elasticity and shape memory effects. However, clinical applications request a deeper knowledge of these properties in order to allow the professional to use them in a rational manner. In addition, the necessary information regarding each alloy often does not correspond to the information given by the manufacturer. Many alloys called "superelastic" do not present this effect; they just behave as less stiff alloys, with a larger springback if compared to the stainless steel wires. CONCLUSIONS: Laboratory tests are the only means to observe the real behavior of these materials, including temperature transition range (TTR and applied tensions. However, it is also possible to determine in which TTR these alloys change the crystalline structure.

Tungsen--nickel--cobalt alloy and method of producing same

International Nuclear Information System (INIS)

Dickinson, J.M.; Riley, R.E.

1977-01-01

An improved tungsten alloy having a tungsten content of approximately 95 weight percent, a nickel content of about 3 weight percent, and the balance being cobalt of about 2 weight percent is described. A method for producing this tungsten--nickel--cobalt alloy is further described and comprises coating the tungsten particles with a nickel--cobalt alloy, pressing the coated particles into a compact shape, heating the compact in hydrogen to a temperature in the range of 1400 0 C and holding at this elevated temperature for a period of about 2 hours, increasing this elevated temperature to about 1500 0 C and holding for 1 hour at this temperature, cooling to about 1200 0 C and replacing the hydrogen atmosphere with an inert argon atmosphere while maintaining this elevated temperature for a period of about 1 / 2 hour, and cooling the resulting alloy to room temperature in this argon atmosphere

Characterization of zinc–nickel alloy electrodeposits obtained from ...

Indian Academy of Sciences (India)

Zinc alloy offers superior sacrificial protection to steel as the alloy dissolves more slowly than pure zinc. The degree of protection and the rate of dissolution depend on the alloying metal and its composition. Zinc-nickel alloy may also serve as at less toxic substitute for cadmium. In this paper the physico-chemical ...

Iron-nickel-chromium alloys

International Nuclear Information System (INIS)

Karenko, M.K.

1981-01-01

A specification is given for iron-nickel-chromium age-hardenable alloys suitable for use in fast breeder reactor ducts and cladding, which utilize the gamma-double prime strengthening phase and are characterized in having a delta or eta phase distributed at or near grain boundaries. A range of compositions is given. (author)

Study of point defect clustering in electron and ion irradiated zirconium alloys

International Nuclear Information System (INIS)

Hellio, C.; Boulanger, L.

1986-09-01

Dislocation loops created by 500 keV Zr + ions and 1 MeV electrons in zirconium have a/3 type Burgers vectors, and in ion irradiated samples, loops lie preferentially on planes close to (1010). From in-situ observations of loop growth under 1 MeV electron irradiation in zirconium and dilute Zr (Nb,O) alloys, a strong increase of the vacancy migration energy with oxygen concentration was observed, from 0.72 eV for pure zirconium to 1.7 eV for Zr and Zr-1% Nb doped with 1800 ppm weight oxygen, indicating large trapping of vacancies by O single interstitials or clusters

Electrodeposition behavior of nickel and nickel-zinc alloys from the zinc chloride-1-ethyl-3-methylimidazolium chloride low temperature molten salt

International Nuclear Information System (INIS)

Gou Shiping; Sun, I.-W.

2008-01-01

The electrodeposition of nickel and nickel-zinc alloys was investigated at polycrystalline tungsten electrode in the zinc chloride-1-ethyl-3-methylimidazolium chloride molten salt. Although nickel(II) chloride dissolved easily into the pure chloride-rich 1-ethyl-3-methylimidazolium chloride ionic melt, metallic nickel could not be obtained by electrochemical reduction of this solution. The addition of zinc chloride to this solution shifted the reduction of nickel(II) to more positive potential making the electrodeposition of nickel possible. The electrodeposition of nickel, however, requires an overpotential driven nucleation process. Dense and compact nickel deposits with good adherence could be prepared by controlling the deposition potential. X-ray powder diffraction measurements indicated the presence of crystalline nickel deposits. Non-anomalous electrodeposition of nickel-zinc alloys was achieved through the underpotential deposition of zinc on the deposited nickel at a potential more negative than that of the deposition of nickel. X-ray powder diffraction and energy-dispersive spectrometry measurements of the electrodeposits indicated that the composition and the phase types of the nickel-zinc alloys are dependent on the deposition potential. For the Ni-Zn alloy deposits prepared by underpotential deposition of Zn on Ni, the Zn content in the Ni-Zn was always less than 50 atom%

Hot corrosion studies on nickel-based alloys containing silicon

International Nuclear Information System (INIS)

Kerr, T.W.; Simkovich, G.

1976-01-01

Alloys of Ni--Cr, Ni--Si and Ni--Cr--Si were oxidized and ''hot corroded'' in pure oxygen at 1000 0 C. In the oxidation experiments it was found that small amounts of either chromium or silicon in nickel increased the oxidation rates in comparison to pure nickel in accord with Wagner's parabolic oxidation theory. At high concentrations of the alloying elements the oxidation rates decreased due to the formation of oxide phases other than nickel oxide in the scale. Hot corrosion experiments were conducted on both binary and ternary alloys by oxidizing samples coated with 1.0 mg/cm 2 of Na 2 SO 4 in oxygen at 1000 0 C. In general it was found that high chromium and high silicon alloys displayed excellent resistance to the hot corrosion process gaining or losing less than 0.5 mg/cm 2 in 1800 min at temperature. Microprobe and x-ray diffraction studies of the alloy and the scale indicate that amorphous SiO 2 probably formed to aid in retarding both the oxidation and the hot corrosion process

Influence of irradiation and radiolysis on the corrosion rates and mechanisms of zirconium alloys

International Nuclear Information System (INIS)

Verlet, Romain

2015-01-01

The nuclear fuel of pressurized water reactors (PWR) in the form of uranium oxide UO 2 pellets (or MOX) is confined in a zirconium alloy cladding. This cladding is very important because it represents the first containment barrier against the release of fission products generated by the nuclear reaction to the external environment. Corrosion by the primary medium of zirconium alloys, particularly the Zircaloy-4, is one of the factors limiting the reactor residence time of the fuel rods (UO 2 pellets + cladding). To optimize core management and to extend the lifetime of the fuel rods in reactor, new alloys based on zirconium-niobium (M5) have been developed. However, the corrosion mechanisms of these are not completely understood because of the complexity of these materials, corrosion environment and the presence of radiation from the nuclear fuel. Therefore, this thesis specifically addresses the effects of radiolysis and defects induced by irradiation with ions in the matrix metal and the oxide layer on the corrosion rate of Zircaloy-4 and M5. The goal is to separate the influence of radiation damage to the metal, that relating to defects created in the oxide and that linked to radiolysis of the primary medium on the oxidation rate of zirconium alloys in reactor. 1) Regarding effect of irradiation of the metal on the oxidation rate: type dislocation loops appear and increase the oxidation rate of the two alloys. For M5, in addition to the first effect, a precipitation of fines needles of niobium reduced the solid solution of niobium concentration in the metal and ultimately in the oxide, which strongly reduces the oxidation rate of the alloy. 2) Regarding the effect of irradiation of the oxide layer on the oxidation rate: defects generated by the nuclear cascades in the oxide increase the oxidation rate of the two materials. For M5, germination of niobium enriched zones in irradiated oxide also causes a decrease of the niobium concentration in solid solution

History of the development of zirconium alloys for use in nuclear reactors

International Nuclear Information System (INIS)

Rickover, H.G.; Geiger, L.D.; Lustman, B.

1975-01-01

The technical problems and the major decisions made during the early development of zirconium alloys for use in naval reactors are outlined. A summary is given of the development of commercial sources of supply for zirconium and hafnium metal over the period 1950 to 1965, and the problems encountered in obtaining zirconium needed for early naval prototype and shipboard reactors are identified. Steps taken in the Government procurement process are described and statistics on production amounts, prices, and inventory are included. Also included are the technical aspects associated with the development of zirconium for water-cooled nuclear reactors, beginning in early 1949 when the Bettis Atomic Power Laboratory was established as a part of the Naval Reactors Program. While in the course of the next 25 years, small-scale investigations were performed on other potential core structural materials such as stainless steel, niobium, aluminum, and beryllium, the pressure for continual development, improvement, and application of zirconium was predominant and unrelenting. (U.S.)

Determination of the gaseous hydrogen ductile-brittle transition in copper-nickel alloys

Science.gov (United States)

Parr, R. A.; Johnston, M. H.; Davis, J. H.; Oh, T. K.

1985-01-01

A series of copper-nickel alloys were fabricated, notched tensile specimens machined for each alloy, and the specimens tested in 34.5 MPa hydrogen and in air. A notched tensile ratio was determined for each alloy and the hydrogen environment embrittlement (HEE) determined for the alloys of 47.7 weight percent nickel to 73.5 weight percent nickel. Stacking fault probability and stacking fault energies were determined for each alloy using the x ray diffraction line shift and line profiles technique. Hydrogen environment embrittlement was determined to be influenced by stacking fault energies; however, the correlation is believed to be indirect and only partially responsible for the HEE behavior of these alloys.

Long-life fatigue test results for two nickel-base structural alloys

International Nuclear Information System (INIS)

Mowbray, D.F.; Giaquinto, E.V.; Mehringer, F.J.

1978-11-01

The results are reported of fatigue tests on two nickel--base alloys, hot-cold-worked and stress-relieved nickel--chrome--iron Alloy 600 and mill-annealed nickel--chrome--moly--iron Alloy 625 in which S-N data were obtained in the life range of 10 6 to 10 10 cycles. The tests were conducted in air at 600 0 F, in the reversed membrane loading mode, at a frequency of approx. 1850 Hz. An electromagnetic, closed loop servo-controlled machine was built to perform the tests. A description of the machine is given

Experimental and numerical study of the effects of a nanocrystallisation treatment on high-temperature oxidation of a zirconium alloy

International Nuclear Information System (INIS)

Panicaud, B.; Retraint, D.; Grosseau-Poussard, J.-L.; Li, L.; Guérain, M.; Goudeau, P.; Tamura, N.; Kunz, M.

2012-01-01

Highlights: ► SMAT leads to a modification of surface properties of an M5 zirconium alloy (grain size and roughness. ► SMAT induces a change in the oxidation kinetics during high temperature oxidation. ► A diffusion model is able to reproduce kinetics and emphasise the consequences of SMAT on dissolution of oxygen in Zr. - Abstract: In the present work, the effects of a nanocrystallisation treatment on the high-temperature oxidation of a zirconium alloy are investigated. Surface Mechanical Attrition Treatment is a recent process designed to nanocrystallise the surface of materials. The particular effects of this treatment on an M5 zirconium alloy are studied using different experimental techniques at several scales. This material is of considerable interest, especially to the nuclear industry where very stringent conditions apply. High temperature oxidation was performed in order to show the benefits of this type of nanocrystallisation on the corrosion resistance of the alloy concerned. Microstructure development mechanisms, which improve the oxidation resistance of zirconium alloys have been identified during high-temperature corrosion. Those mechanisms have been discussed in further detail in relation to numerical calculations concerning the oxidation kinetics.

Electrochemical formation of uranium-zirconium alloy in LiCl-KCl melts

Energy Technology Data Exchange (ETDEWEB)

Murakami, Tsuyoshi, E-mail: m-tsuyo@criepi.denken.or.j [Central Research Institute of Electric Power Industry (CRIEPI), Komae-shi, Tokyo 201-8511 (Japan); Kato, Tetsuya; Kurata, Masaki [Central Research Institute of Electric Power Industry (CRIEPI), Komae-shi, Tokyo 201-8511 (Japan); Yamana, Hajimu [Research Reactor Institute, Kyoto University, Kumatori-cho, Sennan-gun, Osaka 590-0494 (Japan)

2009-11-15

Since zirconium is considered an electrochemically active species under practical conditions of the electrorefining process, it is crucial to understand the electrochemical behavior of zirconium in LiCl-KCl melts containing actinide ions. In this study, the electrochemical codeposition of uranium and zirconium on a solid cathode was performed. It was found that the delta-(U, Zr) phase, which is the only intermediate phase of the uranium-zirconium binary alloy system, was deposited on a tantalum substrate by potentiostatic electrolysis at -1.60 V (vs. Ag{sup +}/Ag) in LiCl-KCl melts containing 0.13 in mol% UCl{sub 3} and 0.23 in mol% ZrCl{sub 4} at 773 K. To our knowledge, this is the first report on the electrochemical formation of the delta-(U, Zr) phase. The relative partial molar properties of uranium in the delta-(U, Zr) phase were evaluated by measuring the open-circuit-potentials of the electrochemically prepared delta-phase electrode.

Electrochemical formation of uranium-zirconium alloy in LiCl-KCl melts

International Nuclear Information System (INIS)

Murakami, Tsuyoshi; Kato, Tetsuya; Kurata, Masaki; Yamana, Hajimu

2009-01-01

Since zirconium is considered an electrochemically active species under practical conditions of the electrorefining process, it is crucial to understand the electrochemical behavior of zirconium in LiCl-KCl melts containing actinide ions. In this study, the electrochemical codeposition of uranium and zirconium on a solid cathode was performed. It was found that the δ-(U, Zr) phase, which is the only intermediate phase of the uranium-zirconium binary alloy system, was deposited on a tantalum substrate by potentiostatic electrolysis at -1.60 V (vs. Ag + /Ag) in LiCl-KCl melts containing 0.13 in mol% UCl 3 and 0.23 in mol% ZrCl 4 at 773 K. To our knowledge, this is the first report on the electrochemical formation of the δ-(U, Zr) phase. The relative partial molar properties of uranium in the δ-(U, Zr) phase were evaluated by measuring the open-circuit-potentials of the electrochemically prepared δ-phase electrode.

Heat treatment of nickel alloys

International Nuclear Information System (INIS)

Smith, D.F. Jr.; Clatworthy, E.F.

1975-01-01

A heat treating process is described that can be used to produce desired combinations of strength, ductility, and fabricability characteristics in heat resistant age-hardenable alloys having precipitation-hardening amounts of niobium, titanium, and/or tantalum in a nickel-containing matrix. (U.S.)

Young's modulus of crystal bar zirconium and zirconium alloys (zircaloy-2, zircaloy-4, zirconium-2.5wt% niobium) to 1000 K

International Nuclear Information System (INIS)

Rosinger, H.E.; Ritchie, I.G.; Shillinglaw, A.J.

1975-09-01

This report contains experimentally determined data on the dynamic elastic moduli of zircaloy-2, zircaloy-4, zirconium-2.5wt% niobium and Marz grade crystal bar zirconium. Data on both the dynamic Young's moduli and shear moduli of the alloys have been measured at room temperature and Young's modulus as a function of temperature has been determined over the temperature range 300 K to 1000 K. In every case, Young's modulus decreases linearly with increasing temperature and is expressed by an empirical equation fitted to the data. Differences in Young's modulus values determined from specimens with longitudinal axes parallel and perpendicular to the rolling direction are small, as are the differences between Young's moduli determined from strip, bar stock and fuel sheathing. (author)

Composition and Performance of Nanostructured Zirconium Titanium Conversion Coating on Aluminum-Magnesium Alloys

Directory of Open Access Journals (Sweden)

Sheng-xue Yu

2013-01-01

Full Text Available Nanostructured conversion coating of Al-Mg alloy was obtained via the surface treatment with zirconium titanium salt solution at 25°C for 10 min. The zirconium titanium salt solution is composed of tannic acid 1.00 g·L−1, K2ZrF6 0.75 g·L−1, NaF 1.25 g·L−1, MgSO4 1.0 g/L, and tetra-n-butyl titanate (TBT 0.08 g·L−1. X-ray diffraction (XRD, X-ray photoelectron spectroscopy (XPS, and Fourier transform infrared spectrum (FT-IR were used to characterize the composition and structure of the obtained conversion coating. The morphology of the conversion coating was obtained by atomic force microscopy (AFM and scanning electron microscopy (SEM. Results exhibit that the zirconium titanium salt conversion coating of Al-Mg alloy contains Ti, Zr, Al, F, O, Mg, C, Na, and so on. The conversion coating with nm level thickness is smooth, uniform, and compact. Corrosion resistance of conversion coating was evaluated in the 3.5 wt.% NaCl electrolyte through polarization curves and electrochemical impedance spectrum (EIS. Self-corrosion current density on the nanostructured conversion coating of Al-Mg alloy is 9.7×10-8A·cm-2, which is only 2% of that on the untreated aluminum-magnesium alloy. This result indicates that the corrosion resistance of the conversion coating is improved markedly after chemical conversion treatment.

Contribution to the understanding of zirconium alloy deformation under irradiation at high doses

International Nuclear Information System (INIS)

Gharbi, Nesrine

2015-01-01

The growth of zirconium alloy tubes of PWR fuel assemblies is the result of two phenomena: axial irradiation creep and stress 'free' growth which is correlated to the formation of c-loops at high irradiation doses. This PhD work aims at investigating the coupling between these two phenomena through a fine Transmission Electron Microscopy analysis of the effect of a macroscopic applied stress on the c-loop microstructure. 600 keV Zr + ion irradiations were performed at 300 C on two recrystallized zirconium alloys: Zircaloy-4 and M5. Thanks to a device specifically designed, different tensile or compressive stress levels were applied under ion irradiation. The microstructural observations have shown that the c-loop density reduces in grains oriented with the c-axis close to the direction of the applied tensile stress or far from the direction of the applied compressive stress, which is in good agreement with the SIPA mechanism. Nevertheless, the examination of a large number of grains has revealed dispersion from grain to grain. This dispersion, which could be explained by the intergranular heterogeneities, reduces the magnitude of the stress effect on c-loop microstructure. In parallel to this experimental study, a cluster dynamics model has been able to describe the evolution under irradiation of zirconium and Zircaloy-4 microstructure and to assess the effect of stress on c-loop microstructure. On the macroscopic scale, a physical model was also developed to predict the irradiation growth and creep behaviour of zirconium alloy tubes. (author) [fr

Influence of alloying elements on the dislocation loops created by Zr+ ion irradiation in alpha-zirconium

International Nuclear Information System (INIS)

Hellio, C.; Novion, C.H. de; Boulanger, L.

1987-01-01

Pure zirconium and four (annealed) α - zirconium based alloys (Zr-1760 ppm weight 0, Zr - 1% Nb - 430 ppm 0, Zr-1% Nb-1800 ppm 0, zircaloy 4) have been studied by transmission electron microscopy after 500 keV Zr + ion or 1 MeV electron irradiation performed at high temperature. Type of burgers vectors of the dislocation loops are given; in the case of electron irradiated Zr-1760 ppm 0, the larger loops were found of interstitial type. Alloying elements increase the loop density. The kinetic of loop growth was observed in-situ during 1 MeV electron irradiation between 400 and 700 0 C: oxygen was found to reduce considerably the growth speed of loops. In-situ annealing at 450 or 500 0 C after ion irradiation led to a large coalescence of loops in the case of pure zirconium, but modified only slightly the defect structure of the alloys

Surface treatment for hydrogen storage alloy of nickel/metal hydride battery

Energy Technology Data Exchange (ETDEWEB)

Wu, M.-S.; Wu, H.-R.; Wang, Y.-Y.; Wan, C.-C. [National Tsing Hua Univ., Hsinchu (Taiwan). Dept. of Chemical Engineering

2000-04-28

The electrochemical performance of AB{sub 2}-type (Ti{sub 0.35}Zr{sub 0.65}Ni{sub 1.2}V{sub 0.6}Mn{sub 0.2}Cr{sub 0.2}) and AB{sub 5}-type (MmB{sub 4.3}(Al{sub 0.3}Mn{sub 0.4}){sub 0.5}) hydrogen storage alloys modified by hot KOH etching and electroless nickel coating has been investigated. It is found that the alloy modified with hot KOH solution shows quick activation but at the expense of cycle-life stability. The alloy coated with nickel was effectively improved in both cycle-life stability and discharge capacity. Both the exchange and limiting current densities were increased by modifying the alloys by hot KOH solution dipping or electroless nickel coating as compared with untreated alloy electrode. The electrode with higher exchange current density and limiting current density leads to increased high-rate dischargeability. A duplex surface modified alloy (i.e., alloy first treated with hot KOH solution and then coated with nickel) has been developed, which performs satisfactorily with respect to both quick activation and long cycle life. In addition, the high-rate dischargeability for the electrode with duplex surface modification is superior to that of electrode solely treated with KOH etching or Ni plating. (orig.)

Nickel-based materials and high-alloy, special stainless steels. 2. new rev. and enl. ed.

International Nuclear Information System (INIS)

Heubner, U.; Brill, U.; Hoffmann, T.; Jasner, M.; Kirchheiner, R.; Koecher, R.; Richter, H.; Rockel, M.; White, F.

1993-01-01

The book is intended as a source of information on nickel-based materials and special stainless steels and apart from the up-to-date materials data presents information on recent developments and knowledge gained, so that it may be a valuable aid to materials engineers looking for cost-effective resolutions of their materials problems in the chemical process industry, power plant operation, and high-temperature applications. The book presents eight individual contributions entitled as follows: (1) Nickel-base alloys and high-alloy, special stainless steels. - Materials survey and data sheets (Ulrich Heubner). (2) Corrosion of nickel-base alloys and special stainless steels (Manfred Rockel). (3) Welding of nickel-base alloys and high-alloy, special stainless steels (Theo Hoffmann). (4) High-temperature resistant materials (Ulrich Brill). (5) Application and processing of nickel-base materials in the chemical process industry and in pollution abatement equipment (Reiner Koecher). (6) Selected examples of applications of nickel-base materials in chemical plant (Manfred Jasner, Frederick White). (7) Applications of nickel-base alloys and special stainless steels in power plant. (8) The use of nickel-base alloys and stainless steels in pollution abatement processes (R. Kirchheiner). (orig./MM). 151 figs., 226 refs [de

Modelling of zirconium alloys corrosion in LWRs

International Nuclear Information System (INIS)

Kritskij, V.G.; Berezina, I.G.; Kritskij, A.V.; Stjagkin, P.S.

1999-01-01

Chemical parameters, that exerted effect on Zr+1%Nb alloy corrosion and deserved consideration during reactor operation, were defined and a model was developed to describe the influence of physical and chemical parameters on zirconium alloys corrosion in nuclear power plants. The model is based on the correlation between the zirconium oxide solubility in high-temperature water under the influence of the chemical parameters and the measured values of fuel cladding corrosion under LWR conditions. The intensity of fuel cladding corrosion in the primary circuits depends on the coolant water quality, growth of iron oxide deposits and vaporization portion. Mathematically, the oxidation rate can be expressed as a sum of heat and radiation components. The temperature dependence on the oxidation rate can be described by the Arrenius equation. The radiation component of Zr uniform corrosion equation is a function of several factors such as neutron fluency, the temperature the metallurgical composition and et. We assume that the main factor is the changing of water chemistry and the H 2 O 2 concentration play the determinative role. Probably, the influence of H 2 O 2 is based on the formation of unstable compound ZrO 3 ·nH 2 O and Zr(OH) 4 with high solubility. The validity of the used formulae was confirmed by corrosion measurements on WWER and RBMK fuel cladding. The model can be applied for calculating the reliability of nuclear fuel operation. (author)

Phosphorus effect on structure and physical properties of iron-nickel alloys

International Nuclear Information System (INIS)

Berseneva, F.N.; Kalinin, V.M.; Rybalko, O.F.

1982-01-01

The structure and properties of iron-nickel alloys (30-50 % Ni) containing from 0.02 to 0.5 wt. % P have been investigated. It has been found that phosphorus solubility in iron-nickel alloys at most purified from impurities exceeds limiting solubility values usually observed for commercial alloys. Phosphide eutectics precipitation over the grain boundaries of studied alloys occurs but with phosphorus content equal 0.45 wt. %. The 0.4 wt. % P addition in invar alloys increases saturation magnetization and the Curie point and leads to a more homogeneous structure

Special features of nickel-molybdenum alloy electrodeposition onto screen-type cathodes

International Nuclear Information System (INIS)

Aleksandrova, G.S.; Varypaev, V.N.

1982-01-01

Electrolytic nickel-molybdenum alloy, which has a rather low hydrogen overpotential and high corrosion resistance, is of interest as cathode material in industrial electrolysis. Screen-type electrodes with a nickel-molybdenum coating can be used as nonconsumable cathodes in water-activated magnesium-alloy batteries

Enhanced low-temperature oxidation of zirconium alloys under irradiation

International Nuclear Information System (INIS)

Cox, B.; Fidleris, V.

1989-01-01

The linear growth of relatively thick (>300 nm) interference-colored oxide films on zirconium alloy specimens exposed in the Advanced Test Reactor (ATR) coolant at ≤55 o C was unexpected. Initial ideas were that this was a photoconduction effect. Experiments to study photoconduction in thin anodic zirconium oxide (ZrO 2 ) films in the laboratory were initiated to provide background data. It was found that, in the laboratory, provided a high electric field was maintained across the oxide during ultraviolet (UV) irradiation, enhanced growth of oxide occurred in the irradiated area. Similarly enhanced growth could be obtained on thin thermally formed oxide films that were immersed in an electrolyte with a high electric field superimposed. This enhanced growth was found to be caused by the development of porosity in the barrier oxide layer by an enhanced local dissolution and reprecipitation process during UV irradiation. Similar porosity was observed in the oxide films on the ATR specimens. Since it is not thought that a high electric field could have been present in this instance, localized dissolution of fast-neutron primary recoil tracks may be the operative mechanism. In all instances, the specimens attempt to maintain the normal barrier-layer oxide thickness, which causes the additional oxide growth. Similar mechanisms may have operated during the formation of thick loosely adherent, porous oxides in homogeneous reactor solutions under irradiation, and may be the cause of enhanced oxidation of zirconium alloys in high-temperature water-cooled reactors in some water chemistries. (author)

Nickel-base alloy forgings for advanced high temperature power plants

Energy Technology Data Exchange (ETDEWEB)

Donth, B.; Diwo, A.; Blaes, N.; Bokelmann, D. [Saarschmiede GmbH Freiformschmiede, Voelklingen (Germany)

2008-07-01

The strong efforts to reduce the CO{sub 2} emissions lead to the demand for improved thermal efficiency of coal fired power plants. An increased thermal efficiency can be realised by higher steam temperatures and pressures in the boiler and the turbine. The European development aims for steam temperatures of 700 C which requires the development and use of new materials and also associated process technology for large components. Temperatures of 700 C and above are too high for the application of ferritic steels and therefore only Nickel-Base Alloys can fulfill the required material properties. In particular the Nickel-Base Alloy A617 is the most candidate alloy on which was focused the investigation and development in several German and European programs during the last 10 years. The goal is to verify and improve the attainable material properties and ultrasonic detectability of large Alloy 617 forgings for turbine rotors and boiler parts. For many years Saarschmiede has been manufacturing nickel and cobalt alloys and is participating the research programs by developing the manufacturing routes for large turbine rotor forgings up to a maximum diameter of 1000 mm as well as for forged tubes and valve parts for the boiler side. The experiences in manufacturing and testing of very large forgings made from nickel base alloys for 700 C steam power plants are reported. (orig.)

Influence of hydratation on the characteristics of zirconium alloys oxide layers

Czech Academy of Sciences Publication Activity Database

Gosmanová, G.; Kraus, I.; Kolega, M.; Vrtílková, V.; Weishauptová, Zuzana

2008-01-01

Roč. 54, č. 1 (2008), s. 1576-1580 ISSN 1210-0471 R&D Projects: GA ČR GA106/04/0043 Institutional research plan: CEZ:AV0Z30460519 Keywords : zirconium alloys * corrosion layer * hydrated ZrO2 Subject RIV: JF - Nuclear Energetics

Effects of deposition temperature on electrodeposition of zinc–nickel alloy coatings

International Nuclear Information System (INIS)

Qiao, Xiaoping; Li, Helin; Zhao, Wenzhen; Li, Dejun

2013-01-01

Highlights: ► Both normal and anomalous deposition can be realized by changing bath temperature. ► The Ni content in Zn–Ni alloy deposit increases sharply as temperature reach 60 °C. ► The abrupt change in coating composition is caused by the shift of cathodic potential. ► The deposition temperature has great effect on microstructure of Zn–Ni alloy deposit. -- Abstract: Zinc–nickel alloy coatings were electrodeposited on carbon steel substrates from the ammonium chloride bath at different temperatures. The composition, phase structure and morphology of these coatings were analyzed by energy dispersive spectrometer, X-ray diffractometer and scanning electron microscopy respectively. Chronopotentiometry and potentiostatic methods were also employed to analyze the possible causes of the composition and structure changes induced by deposition temperature. It has been shown that both normal and anomalous co-deposition of zinc and nickel could be realized by changing deposition temperature under galvanostatic conditions. The abrupt changes in the composition and phase structure of the zinc–nickel alloy coatings were observed when deposition temperature reached 60 °C. The sharply decrease of current efficiency for zinc–nickel co-deposition was also observed when deposition temperature is higher than 40 °C. Analysis of the partial current densities showed that the decrease of current efficiency with the rise of deposition temperature was due to the enhancement of the hydrogen evolution. It was also confirmed that the ennoblement of cathodic potential was the cause for the increase of nickel content in zinc–nickel alloy coatings as a result of deposition temperature rise. The good zinc–nickel alloy coatings with compact morphology and single γ phase could be obtained when the deposition temperature was fixed at 30–40 °C

Investigation into cathode polarization during deposition of rhodium-nickel and rhodium-indium alloys

International Nuclear Information System (INIS)

Evdokimova, N.V.; Byacheslavov, P.M.; Lokshtanova, O.G.

1979-01-01

The results of kinetic regularities experimental investigations during electrodeposition of rhodium-nickel and rhonium-indium alloys are presented. Methods of general and partial polarization curves have been used to show the nature of polarization during the rhonium-nickel and rhodium-indium alloys deposition. It is shown that indium into the rhodium-indium alloy and nickel into the rhodium-nickel alloy deposit with great depolarization ( PHIsub(In)sup(0)=-0.33B, PHIsub(Ni)sup(0)=-0.23B). Indium and nickel in pure form do not deposit from the electrolytes of the given composition (H 2 SO 4 - 50 g/l, HNH 2 SO 3 -10 g/l). The recalculation of partial polarization curve of indium precipitation into the rhodium-indium alloy in the mixed kinetics coordinates gives a straight line with 40 mV inclination angle. This corresponds to the delayed stage of the second electron addition with the imposition of diffusion limitations

Evaluation and comparison of shear bond strength of porcelain to a beryllium-free alloy of nickel-chromium, nickel and beryllium free alloy of cobalt-chromium, and titanium: An in vitro study

Directory of Open Access Journals (Sweden)

Ananya Singh

2017-01-01

Conclusion: It could be concluded that newer nickel and beryllium free Co-Cr alloys and titanium alloys with improved strength to weight ratio could prove to be good alternatives to the conventional nickel-based alloys when biocompatibility was a concern.

Corrosion Characterization in Nickel Plated 110 ksi Low Alloy Steel and Incoloy 925: An Experimental Case Study

Science.gov (United States)

Thomas, Kiran; Vincent, S.; Barbadikar, Dipika; Kumar, Shresh; Anwar, Rebin; Fernandes, Nevil

2018-04-01

Incoloy 925 is an age hardenable Nickel-Iron-Chromium alloy with the addition of Molybdenum, Copper, Titanium and Aluminium used in many applications in oil and gas industry. Nickel alloys are preferred mostly in corrosive environments where there is high concentration of H2S, CO2, chlorides and free Sulphur as sufficient nickel content provides protection against chloride-ion stress-corrosion cracking. But unfortunately, Nickel alloys are very expensive. Plating an alloy steel part with nickel would cost much lesser than a part make of nickel alloy for large quantities. A brief study will be carried out to compare the performance of nickel plated alloy steel with that of an Incoloy 925 part by conducting corrosion tests. Tests will be carried out using different coating thicknesses of Nickel on low alloy steel in 0.1 M NaCl solution and results will be verified. From the test results we can confirm that Nickel plated low alloy steel is found to exhibit fairly good corrosion in comparison with Incoloy 925 and thus can be an excellent candidate to replace Incoloy materials.

An overview of advanced high-strength nickel-base alloys for LWR applications

International Nuclear Information System (INIS)

Prybylowski, J.; Ballinger, R.G.

1989-01-01

This paper reviews our current understanding of the behavior of high strength nickel base alloys used in light water reactor (LWR) applications. Emphasis is placed on understanding the fundamental mechanisms controlling crack propagation in these environments. To provide a foundation for this survey, general mechanisms of stress corrosion cracking and hydrogen embrittlement are first reviewed. The behavior of high strength nickel base alloys in LWR environments, as well as in other relevant environments is then reviewed. Suggested mechanisms of crack propagation are discussed. Alternate alloys and microstructural modifications that may result in improved behavior are presented. It is now clear that, at temperatures near 100C, alloy X-750, the predominant high strength nickel base alloy used today in LWR applications, is susceptible to hydrogen embrittlement. A review of published data from hydrogen embrittlement studies of nickel base superalloys during electrolytic charging and in hydrogen sulfide/brine solutions suggests that other nickel base superalloys are available possessing resistance to hydrogen embrittlement superior to that of alloy X-750. Available results of tests in gaseous hydrogen suggest that reduced grain boundary precipitation and a fine distribution of intragranular precipitates that act as irreversible hydrogen traps is the optimum microstructure for hydrogen embrittlement resistance. 42 refs., 2 figs., 5 tabs

Study of the oxidation kinetics of the nickel-molybdenum alloy

International Nuclear Information System (INIS)

Gouillon, Marie-Josephe

1974-01-01

This research thesis reports the study of the oxidation of a nickel-molybdenum alloy in the high-nickel-content part of this alloy. After a bibliographical study on the both metals, the author proposes a physical model based on observed phenomena and based on experimental results. Based on a thermodynamic study, the author compares the stability of the different oxides which may be formed, and reports a prediction of oxides obtained on the alloy during oxidation. Qualitative and quantitative studies have been performed by scanning electron microscopy coupled with electronic microprobe analysis to investigate morphological characteristics on oxidation films. A kinetic study by thermogravimetry shows a decrease of the alloy oxidation rate with respect to that of pure nickel at temperatures lower than 800 degrees C. This result is interpreted by the intervention of two opposed diffusion phenomena which act against each other [fr

Microstructural aspects of the oxidation of zirconium alloys

International Nuclear Information System (INIS)

Proff, Ch.

2011-01-01

This thesis is focused on the microstructural characterisation of precipitates in the oxide of binary zirconium alloys (1 wt.% Fe, Cr or Ni or 0.6 wt.% Nb) under different oxidation conditions at 415 C. The samples were oxidised in autoclave in air and steam and in an environmental scanning electron microscope in water vapour. The microstructural evolution of the precipitates during oxidation was characterised using electron microscopy. The findings from the analysis are the following: -Two types of oxidation behaviour are observed for precipitates. -Pilling Bedworth ratio of precipitates is higher than that of the zirconium matrix. -Formation of pure iron oxide crystals on the surface for iron bearing precipitates close to or at the surface. From these observations it is concluded that the precipitate oxidation behaviour can be correlated to precipitate composition and oxidation tendency of the elements in the precipitates. Iron exhibits clearly different behaviour. (author)

Application of FEM analytical method for hydrogen migration behaviour in Zirconium alloys

Energy Technology Data Exchange (ETDEWEB)

Arioka, K; Ohta, H [Takasago Research and Development Center, Mitsubishi Heavy Industries Ltd, Hyogo-ken (Japan)

1997-02-01

It is well recognized that the hydriding behaviours of Zirconium alloys are very significant problems as a safety issues. Also, it is well known that the diffusion of hydrogen in Zirconium alloys are affected not only by concentration but also temperature gradient. But in actual component, especially heat transfer tube such as fuel rod, we can not avoid the temperature gradient in some degree. So, it is very useful to develop the computer code which can analyze the hydrogen diffusion and precipitation behaviours under temperature gradient as a function of the structure of fuel rod. For this objective, we have developed the computer code for hydrogen migration behaviour using FEM analytical methods. So, following items are presented and discussed. Analytical method and conditions; correlation between the computed and test results; application to designing studies. (author). 8 refs, 4 figs, 2 tabs.

Round robin test for zirconium alloys in 400 deg C steam: results from EDF; Essais interlaboratoires de corrosion generalisee en milieu vapeur a 400 deg C d`alliages de zirconium: resultats d`EDF

Energy Technology Data Exchange (ETDEWEB)

Blat, M.

1994-01-01

The EDF Material Studies Branch has participated in the Round Robin program of uniform corrosion on zirconium alloys. The objectives of these Round Robin corrosion tests are to generate new uniform corrosion weight gain date utilizing modern zirconium alloy products and to improve the International and ASTM standards. (author). 2 tabs., 7 appendix., 2 refs.

Aluminium alloys containing iron and nickel

International Nuclear Information System (INIS)

Coriou, H.; Fournier, R.; Grall, L.; Hure, J.; Herenguel, J.; Lelong, P.

1958-01-01

The first part of this report addresses mechanism, kinetics and structure factors of aluminium alloys containing iron and nickel in water and high temperature steam. The studied alloys contain from 0.3 to 0.7 per cent of iron, and 0.2 to 1.0 per cent of nickel. Corrosion resistance and corrosion structure have been studied. The experimental installation, process and samples are presented. Corrosion structures in water at 350 C are identified and discussed (structure of corrosion products, structure of metal-oxide interface), and then in steam at different temperatures (350-395 C). Corrosion kinetics is experimentally studied (weight variation in time) in water at 350 C and in steam at different temperatures. Reactions occurring at over-heated steam (more than 400 C) are studied, and the case of welded alloys is also addressed. The second part addresses the metallurgical mechanism and processes influencing aluminium alloy resistance to corrosion by high temperature water as it appeared that separated phases protect the solid solution through a neighbourhood action. In order to avoid deep local corrosions, it seems necessary to multiply protective phases in an as uniform as possible way. Some processes enabling this result are described. They belong to conventional metallurgy or to powder metallurgy (with sintering and extrusion)

Method for inhibiting corrosion of nickel-containing alloys

Science.gov (United States)

DeVan, J.H.; Selle, J.E.

Nickel-containing alloys are protected against corrosion by contacting the alloy with a molten alkali metal having dissolved therein aluminum, silicon or manganese to cause the formation of a corrosion-resistant intermetallic layer. Components can be protected by applying the coating after an apparatus is assembled.

Characterization of zirconium alloy oxidation films by alternating current impedance

International Nuclear Information System (INIS)

Rosecrans, P.M.

1983-11-01

Kinetics of zirocnium alloy oxidation are highly nonlinear. The results of electrochemical measurements and electron microscopy support the existence of porosity in oxide films formed on zirconium alloys in high temperature aqueous environments. Analytical treatment is presented relating oxidation kinetics to the thickness and distribution of nonporous elements within the oxide. This analysis illustrates that both the level and distribution of porosity within the oxide factor into oxidation kinetics. The barrier layer model can provide a basis for predicting the effect of environmental changes on oxidation rate. In addition, it demonstrates the need for further research into porosity generation mechanisms in oxide films

Phase composition and properties of rapidly cooled aluminium-zirconium-chromium alloys

International Nuclear Information System (INIS)

Sokolovskaya, E.M.; Badalova, L.M.; Podd''yakova, E.I.; Kazakova, E.F.; Loboda, T.P.; Gribanov, A.V.

1989-01-01

Using the methods of physicochemical analysis the interaction of aluminium with zirconium and chromium is studied. Polythermal cross sections between Al 3 -Zr-Al 7 Cr and radial polythermal cross section from aluminium-rich corner with the ratio of components Zr:Cr=5:7 by mass are constructed. The effect of zirconium and chromium content on electrochemical characteristics of aluminium-base rapidly quenching alloys in systems Al-Cr, Al-Zr, Al-Cr-Zr. An increase in chromium concentration in oversaturated solid solution of Al-Cr system expands considerably the range of passive state. When Al 7 Cr phase appears the range of passive stae vanishes

Critical assessment of finite element analysis applied to metal–oxide interface roughness in oxidising zirconium alloys

Energy Technology Data Exchange (ETDEWEB)

Platt, P., E-mail: Philip.Platt@manchester.ac.uk [University of Manchester, School of Materials, Materials Performance Centre, Manchester M13 9PL (United Kingdom); Frankel, P. [University of Manchester, School of Materials, Materials Performance Centre, Manchester M13 9PL (United Kingdom); Gass, M. [AMEC, Walton House, Faraday Street, Birchwood Park, Risley, Warrington WA3 6GA (United Kingdom); Preuss, M. [University of Manchester, School of Materials, Materials Performance Centre, Manchester M13 9PL (United Kingdom)

2015-09-15

As a nuclear fuel cladding material, zirconium alloys act as a barrier between the fuel and pressurised steam or lithiated water environment. Controlling degradation mechanisms such as oxidation is essential to extending the in-service lifetime of the fuel. At temperatures of ∼360 °C zirconium alloys are known to exhibit cyclical, approximately cubic corrosion kinetics. With acceleration in the oxidation kinetics occurring every ∼2 μm of oxide growth, and being associated with the formation of a network of lateral cracks. Finite element analysis has been used previously to explain the lateral crack formation by the development of localised out-of-plane tensile stresses at the metal–oxide interface. This work uses the Abaqus finite element code to assess critically current approaches to representing the oxidation of zirconium alloys, with relation to undulations at the metal–oxide interface and localised stress generation. This includes comparison of axisymmetric and 3D quartered modelling approaches, and investigates the effect of interface geometry and plasticity in the metal substrate. Particular focus is placed on the application of the anisotropic strain tensor used to represent the oxidation mechanism, which is typically applied with a fixed coordinate system. Assessment of the impact of the tensor showed that 99% of the localised tensile stresses originated from the out-of-plane component of the strain tensor, rather than the in-plane expansion as was previously thought. Discussion is given to the difficulties associated with this modelling approach and the requirements for future simulations of the oxidation of zirconium alloys.

Method for electrodeposition of nickel--chromium alloys and coating of uranium

International Nuclear Information System (INIS)

Stromatt, R.W.; Lundquist, J.R.

1975-01-01

High-quality electrodeposits of nickel-chromium binary alloys in which the percentage of chromium is controlled can be obtained by the addition of a complexing agent such as ethylenediaminetetraacetic disodium salt to the plating solution. The nickel-chromium alloys were found to provide an excellent hydrogen barrier for the protection of uranium fuel elements. (U.S.)

Effects of Oxidation and fractal surface roughness on the wettability and critical heat flux of glass-peened zirconium alloy tubes

International Nuclear Information System (INIS)

Fong, R.W.L.; Nitheanandan, T.; Bullock, C.D.; Slater, L.F.; McRae, G.A.

2003-05-01

Glass-bead peening the outside surfaces of zirconium alloy tubes has been shown to increase the Critical Heat Flux (CHF) in pool boiling of water. The CHF is found to correlate with the fractal roughness of the metal tube surfaces. In this study on the effect of oxidation on glass-peened surfaces, test measurements for CHF, surface wettability and roughness have been evaluated using various glass-peened and oxidized zirconium alloy tubes. The results show that oxidation changes the solid-liquid contact angle (i.e., decreases wettability of the metal-oxide surface), but does not change the fractal surface roughness, appreciably. Thus, oxidation of the glass-peened surfaces of zirconium alloy tubes is not expected to degrade the CHF enhancement obtained by glass-bead peening. (author)

Fluorimetric determination of uranium in zirconium and zircaloy alloys; Determinacion fluorimetrica de uranio en aleaciones de zirconio y zircaloy

Energy Technology Data Exchange (ETDEWEB)

Acosta L, E [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)

1991-05-15

The objective of this procedure is to determine microquantities of uranium in zirconium and zircaloy alloys. The report also covers the determination of uranium in zirconium alloys and zircaloy in the range from 0.25 to 20 ppm on 1 g of base sample of radioactive material. These limit its can be variable if the size of the used aliquot one is changed for the final determination of uranium. (Author)

Growth and characterization of oxide layers on zirconium alloys

International Nuclear Information System (INIS)

Maroto, A.J.G.; Bordoni, R.; Villegas, M.; Blesa, M.A.; Olmedo, A.M.; Iglesias, A.; Rigotti, G.

1997-01-01

Corrosion behaviour in aqueous media at high temperature of zirconium alloys has been extensively studied in order to elucidate the corrosion mechanism and kinetics. The characterization of the morphology and microstructure of these oxides through the different stages of oxide growth may contribute to understand their corrosion mechanism. Argentina has initiated a research program to correlate long term in and out-reactor corrosion of these alloys. This paper reports a comparative study of out of pile oxidation of Zr-2.5Nb and Zry-4, which are structural materials of in-core components of nuclear power plants. Kinetic data at different temperatures and microstructural characterization of the oxide films are presented. (author). 25 refs, 18 figs, 1 tab

Growth and characterization of oxide layers on zirconium alloys

Energy Technology Data Exchange (ETDEWEB)

Maroto, A J.G.; Bordoni, R; Villegas, M; Blesa, M A; Olmedo, A M; Iglesias, A; Rigotti, G [Comision Nacional de Energia Atomica, Buenos Aires (Argentina)

1997-02-01

Corrosion behaviour in aqueous media at high temperature of zirconium alloys has been extensively studied in order to elucidate the corrosion mechanism and kinetics. The characterization of the morphology and microstructure of these oxides through the different stages of oxide growth may contribute to understand their corrosion mechanism. Argentina has initiated a research program to correlate long term in and out-reactor corrosion of these alloys. This paper reports a comparative study of out of pile oxidation of Zr-2.5Nb and Zry-4, which are structural materials of in-core components of nuclear power plants. Kinetic data at different temperatures and microstructural characterization of the oxide films are presented. (author). 25 refs, 18 figs, 1 tab.

Tube in zirconium base alloy for nuclear fuel assembly and manufacturing process of such a tube

International Nuclear Information System (INIS)

Mardon, J.P.; Senevat, J.; Charquet, D.

1996-01-01

This patent concerns the description and manufacturing guidelines of a zirconium alloy tube for fuel cladding or fuel assembly guiding. The alloy contains (in weight) 0.4 to 0.6% of tin, 0.5 to 0.8% of iron, 0.35 to 0.50% of vanadium and 0.1 to 0.18% of oxygen. The carbon and silicon tenors range from 100 to 180 ppm and from 80 to 120 ppm, respectively. The alloy contains only zirconium, plus inevitable impurities, and is completely recrystallized. Corrosion resistance tests were performed on tubes made of this alloy and compared to corrosion tests performed on zircaloy 4 tubes. These tests show a better corrosion resistance and a lower corrosion kinetics for the new alloy, even in presence of lithium and iodine, and a lower hydridation rate. The mechanical resistance of this alloy is slightly lower than the one of zircaloy 4 but becomes equivalent or slightly better after two irradiation cycles. The ductility remains always equal or better than for zircaloy 4. (J.S.)

Studies on influence of zinc immersion and fluoride on nickel electroplating on magnesium alloy AZ91D

International Nuclear Information System (INIS)

Zhang Ziping; Yu Gang; Ouyang Yuejun; He Xiaomei; Hu Bonian; Zhang Jun; Wu Zhenjun

2009-01-01

The effect of zinc immersion and the role of fluoride in nickel plating bath were mainly investigated in nickel electroplating on magnesium alloy AZ91D. The state of zinc immersion, the composition of zinc film and the role of fluoride in nickel plating bath were explored from the curves of open circuit potential (OCP) and potentiodynamic polarization, the images of scanning electron microscopy (SEM) and the patterns of energy dispersive X-ray (EDX). Results show that the optimum zinc film mixing small amount of Mg(OH) 2 and MgF 2 is obtained by zinc immersion for 30-90 s. The corrosion potential of magnesium alloy substrate attached zinc film will be increased in nickel plating bath and the quantity of MgF 2 sandwiched between magnesium alloy substrate and nickel coating will be reduced, which contributed to produce nickel coating with good performance. Fluoride in nickel plating bath serves as an activator of nickel anodic dissolution and corrosion inhibitor of magnesium alloy substrate. 1.0-1.5 mol dm -3 of F - is the optimum concentration range for dissolving nickel anode and protecting magnesium alloy substrate from over-corrosion in nickel plating bath. The nickel coating with good adhesion and high corrosion resistance on magnesium alloy AZ91D is obtained by the developed process of nickel electroplating. This nickel layer can be used as the rendering coating for further plating on magnesium alloys.

Hot-rolled and cold-finished zirconium and zirconium alloy bars, rod, and wire for nuclear application

International Nuclear Information System (INIS)

Anon.

1981-01-01

The specification covers hot- and cold-finished zirconium alloy bars, rod, and wire, other than those required for reforging, including rounds, squares, and shapes. One unalloyed grade and three alloy grades for use in nuclear applications are described. The products covered include the following sections and sizes: bars, rounds in coils for subsequent reworking (6.4 to 19 mm) and flats (6.4 to 250 mm); rods, rounds in coils for subsequent reworking (6.4 to 19 mm); wire (9.5 mm). The specification covers ordering information, manufacture, condition, chemical requirements, mechanical properties, corrosion properties, permissible variations in dimensions, significance of numerical limits, lot size, special tests, workmanship, finish, inspection, certification, packaging and marking

ZIRCONIUM-CLADDING OF THORIUM

Science.gov (United States)

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)

Mechanisms of irradiation growth of alpha-zirconium alloys

International Nuclear Information System (INIS)

Holt, R.A.

1988-01-01

Experimental observations in the last few years have shown that the range of irradiation growth behaviour of alpha-zirconium alloys is more varied, that a wider variety of sinks must be considered, and that there are more potential sources of anisotropy than was previously recognized. The important new experimental observations which influence our preception of the growth phenomenon in zirconium alloys include the growth of single crystals, accelerating growth in annealed material with the coincident appearance of vacancy loops on the basal planes, the occurrence of 'negative' growth, i.e., contractions along prism directions, the absence of a pronounced effect of grain size on the long term growth rate at low temperatures, and the presence of intergranular constraints prior to irradiation. With the greater complexity of behaviour now being observed, it is necessary to apply new theoretical concepts to assist in understanding growth, e.g., the potential role of anisotropic diffusion in segregation point defects to different sinks and 'growth' caused by the anisotropic relaxation of intergranular constrains. These can be combined with earlier ideas to predict a variety of growth behaviours, including 'negative growth'. Because the most important physical information required for theoretical treatments of growth, i.e, the characteristics of vacancies and self interstitial atoms, are still poorly understood, it is almost impossible to test rigorously any particular theoretical concept and a complete picture of growth has yet to emerge. (orig./MM)

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

International Nuclear Information System (INIS)

Sabol, G.P.; Barry, R.F.

1987-01-01

A process is described for forming seamless tubing of a material selected from zirconium, zirconium alloys, titanium, and titanium alloys, from welded precursor tubing of the material, having a heterogeneous structure resulting from the welding thereof. The process consists of: heating successive axial segments of the welded tubing, completely through the wall thereof, including the weld, to uniformly transform the heterogeneous, as welded, material into the beta phase; quenching the beta phase tubing segments, the heating and quenching effected sufficiently rapid enough to produce a fine sized beta grain structure completely throughout the precursor tubing, including the weld, and to prevent growth of beta grains within the material larger than 200 micrometers in diameter; and subsequently uniformly deforming the quenched precursor tubing by cold reduction steps to produce a seamless tubing of final size and shape

Determination of impurities in uranium--niobium (7.5%)--zirconium (2.5%) alloy

Energy Technology Data Exchange (ETDEWEB)

Arragon, Y

1973-10-01

The determination of 11 impurities in uranium--niobium-- zirconium alloys was studied. Elements of which the alloy is composed are considered and information is given on the determination of niobium by niobic acid precipitation. Selective elimination of the three components is discussed. Two liquid-liquid extractions are used. The nioblum is separated by methylisobutylketone in a hydrochloric --hydrofluoric medium and the zirconium and uranium by tributyl phosphate in a nitric medium. The determination of trace elements using electrochemical methods is discussed. Anodic re-dissolution polarography or square wave polarography enabled six elements (cadmium, copper, lead, zinc, bismuth, and thallium) to be determined in a carbonate medium together with aluminium in tetraethylammonium perchlorate, molybdenum in nitric acid, ammonium nitrate, and tungsten in hydrochloric acid with added double sodium and potassium tartrate. Fluorine was determined using ionometric techniques with a specific electrode and carbon was titrated by conductometry after combustion of the sample in an oxygen current. (auth)

Corrosion Behavior of Nickel-Plated Alloy 600 in High Temperature Water

International Nuclear Information System (INIS)

Kim, Ji Hyun; Hwang, Il Soon

2008-01-01

In this paper, electrochemical and microstructural characteristics of nickel-plated Alloy 600 wee investigated in order to identify the performance of electroless Ni-plating on Alloy 600 in high-temperature aqueous condition with the comparison of electrolytic nickel-plating. For high temperature corrosion test of nickel-plated Alloy 600, specimens were exposed for 770 hours to typical PWR primary water condition. During the test, open circuit potentials (OCP's) of all specimens were measured using a reference electrode. Also, resistance to flow accelerated corrosion (FAC) test was examined in order to check the durability of plated layers in high-velocity flow environment at high temperature. After exposures to high flow rate aqueous condition, the integrity of surfaces was confirmed by using both scanning electron microscopy (SEM) and energy dispersive spectroscopy (EDS). For the field application, a remote process for electroless nickel-plating was demonstrated using a plate specimen with narrow gap on a laboratory scale. Finally, a practical seal design was suggested for more convenient application

Prevention of delayed hydride cracking in zirconium alloys

International Nuclear Information System (INIS)

Cheadle, B.A.; Coleman, C.E.; Ambler, J.F.R.

1987-01-01

Zirconium alloys are susceptible to a mechanism for crack initiation and propagation called delayed hydride cracking. From a review of component failures and experimental results, we have developed the requirements for preventing this cracking. The important parameters for cracking are hydrogen concentration, flaws, and stress; each should be minimized. At the design and construction stages hydrogen pickup has to be controlled, quality assurance needs to be at a high enough level to ensure the absence of flaws, and residual stresses must be eliminated by careful fabrication and heat treatment

Advances in zirconium technology for nuclear reactor application

International Nuclear Information System (INIS)

Ganguly, C.

2002-01-01

Zirconium alloys are extensively used as a material for cladding nuclear fuels and for making core structurals of water-cooled nuclear power reactors all over the world for generation of nearly 16 percent of the worlds electricity. Only four countries in the world, namely France, USA, Russia and India, have large zirconium industry and capability to manufacture reactor grade zirconium sponge, a number of zirconium alloys and a wide variety of structural components for water cooled nuclear reactor. The present paper summarises the status of zirconium technology and highlights the achievement of Nuclear Fuel Complex during the last ten years in developing a wide variety of zirconium alloys and components for water-cooled nuclear power programme

Mechanical and wear properties of pre-alloyed molybdenum P/M steels with nickel addition

Directory of Open Access Journals (Sweden)

Yamanoglu R.

2012-01-01

Full Text Available The aim of this study is to understand the effect of nickel addition on mechanical and wear properties of molybdenum and copper alloyed P/M steel. Specimens with three different nickel contents were pressed under 400 MPa and sintered at 1120ºC for 30 minutes then rapidly cooled. Microstructures and mechanical properties (bending strength, hardness and wear properties of the sintered specimens were investigated in detail. Metallographical investigations showed that the microstructures of consolidated specimens consist of tempered martensite, bainite, retained austenite and pores. It is also reported that the amount of pores varies depending on the nickel concentration of the alloys. Hardness of the alloys increases with increasing nickel content. Specimens containing 2% nickel showed minimum pore quantity and maximum wear resistance. The wear mechanism changed from abrasive wear at low nickel content to adhesive wear at higher nickel content.

Multiaxial ratcheting behavior of zirconium alloy tubes under combined cyclic axial load and internal pressure

Energy Technology Data Exchange (ETDEWEB)

Chen, G.; Zhang, X. [School of Chemical Engineering and Technology, Tianjin University, Tianjin 300072 (China); Xu, D.K. [Environmental Corrosion Center, Institute of Metal Research, Chinese Academy of Sciences, Shenyang 110016 (China); Li, D.H. [Hunan Taohuajiang Nuclear Power Co., Ltd, Yiyang, 413000 (China); Chen, X. [School of Chemical Engineering and Technology, Tianjin University, Tianjin 300072 (China); Zhang, Z., E-mail: zhe.zhang@tju.edu.cn [School of Chemical Engineering and Technology, Tianjin University, Tianjin 300072 (China)

2017-06-15

In this study, a series of uniaxial and multiaxial ratcheting tests were conducted at room temperature on zirconium alloy tubes. The experimental results showed that for uniaxial symmetrical cyclic test, the axial ratcheting strain ɛ{sub x} did not accumulate obviously in initial stage, but gradually increased up to 1% with increasing stress amplitude σ{sub xa}. For multiaxial ratcheting tests, the zirconium alloy tube was highly sensitive to both the axial stress amplitude σ{sub xa} and the internal pressure p{sub i}. The hoop ratcheting strain ɛ{sub θ} increased continuously with the increase of axial stress amplitude, whereas the evolution of axial ratcheting strain ɛ{sub x} was related to the axial stress amplitude. The internal pressure restricted the ratcheting accumulation in the axial direction, but promoted the hoop ratcheting strain on the contrary. The prior loading history greatly restrained the ratcheting behavior of subsequent cycling with a small internal pressure. - Highlights: •Uniaxial and multiaxial ratcheting behavior of the zirconium alloy tubes are investigated at room temperature. •The ratcheting depends greatly on the stress amplitude or internal pressure. •The interaction between the axial and hoop ratcheting mechanisms is greatly dependent on the internal pressure level. •The ratcheting is influenced significantly by the loading history of internal pressure.

Multiaxial ratcheting behavior of zirconium alloy tubes under combined cyclic axial load and internal pressure

International Nuclear Information System (INIS)

Chen, G.; Zhang, X.; Xu, D.K.; Li, D.H.; Chen, X.; Zhang, Z.

2017-01-01

In this study, a series of uniaxial and multiaxial ratcheting tests were conducted at room temperature on zirconium alloy tubes. The experimental results showed that for uniaxial symmetrical cyclic test, the axial ratcheting strain ɛ x did not accumulate obviously in initial stage, but gradually increased up to 1% with increasing stress amplitude σ xa . For multiaxial ratcheting tests, the zirconium alloy tube was highly sensitive to both the axial stress amplitude σ xa and the internal pressure p i . The hoop ratcheting strain ɛ θ increased continuously with the increase of axial stress amplitude, whereas the evolution of axial ratcheting strain ɛ x was related to the axial stress amplitude. The internal pressure restricted the ratcheting accumulation in the axial direction, but promoted the hoop ratcheting strain on the contrary. The prior loading history greatly restrained the ratcheting behavior of subsequent cycling with a small internal pressure. - Highlights: •Uniaxial and multiaxial ratcheting behavior of the zirconium alloy tubes are investigated at room temperature. •The ratcheting depends greatly on the stress amplitude or internal pressure. •The interaction between the axial and hoop ratcheting mechanisms is greatly dependent on the internal pressure level. •The ratcheting is influenced significantly by the loading history of internal pressure.

Nanophase materials produced by physical methods

International Nuclear Information System (INIS)

Noda, Shoji

1992-01-01

A nanophase material is mainly characterized by the component's size and the large interface area. Some nanophase materials are briefly described. Ion implantation and oblique vapor deposition are taken as the methods to provide nanophase materials, and their features are described. These physical methods are non-equilibrium material processes, and the unique nanophase materials are demonstrated to be provided by these methods with little thermodynamic restriction. (author)

High-resolution characterization of oxidation mechanism of zirconium nuclear fuel cladding alloys

International Nuclear Information System (INIS)

Hu, J.; Lozano-Perez, S.; Grovenor, C.

2015-01-01

Full text of publication follows. Zirconium alloys are used extensively as cladding materials in modern light water reactors to separate the uranium dioxide (UO 2 ) fuel rods and the coolant water in order to prevent the escape of radioactive fission products whilst maintaining heat transfer to the coolant. With increasing demand for high burn-up in modern nuclear reactors, environmental degradation of these alloys is now the life limiting factor for fuel assemblies. As part of the MUZIC-2 collaboration studying oxidation and hydrogen pickup in Zr alloys, several high resolution analysis techniques have been used to study the microstructure of a range of commercial and developmental Zr alloys. The sample used for this investigation was prepared from a Westinghouse TM developmental alloy with composition of Zr-0.9Nb-0.01Sn-0.08Fe (wt %) in the recrystallized condition. The sample was oxidised in an autoclave at EDF Energy under simulated PWR water conditions at 360 C. degrees for 360 days. Using Transmission Electron Microscope (TEM), we have studied the development of the equiaxed-columnar-equiaxed grain structure, and observe that the columnar grains are both longer and show a stronger preferred texture in more corrosion-resistant alloys. Fresnel imaging revealed the existence of both parallel interconnected pores and some vertically interconnected pores along the columnar oxide grain boundaries, which become more disconnected near the metal-oxide interface. Electron Energy Loss Spectroscopy (EELS) provided accurate quantitative analysis of the oxygen concentration across the interface, identifying the existence of local regions of stoichiometric ZrO and Zr 3 O 2 with varying thickness. These observations will be discussed in the context of current models for oxidation in zirconium alloys. (authors)

Unloading Effect on Delayed Hydride Cracking in Zirconium Alloys

International Nuclear Information System (INIS)

Kim, Young Suk; Kim, Sung Soo

2010-01-01

It is well-known that a tensile overload retards not only the crack growth rate (CGR) in zirconium alloys during the delayed hydride cracking (DHC) tests but also the fatigue crack growth rate in metals, the cause of which is unclear to date. A considerable decrease in the fatigue crack growth rate due to overload is suggested to occur due either to the crack closure or to compressive stresses or strains arising from unloading of the overload. However, the role of the crack closure or the compressive stress in the crack growth rate remains yet to be understood because of incomplete understanding of crack growth kinetics. The aim of this study is to resolve the effect of unloading on the CGR of zirconium alloys, which comes in last among the unresolved issues as listed above. To this end, the CGRs of the Zr-2.5Nb tubes were determined at a constant temperature under the cyclic load with the load ratio, R changing from 0.13 to 0.66 where the extent of unloading became higher at the lower R. More direct evidence for the effect of unloading after an overload is provided using Simpson's experiment investigating the effect on the CGR of a Zr-2.5Nb tube of the stress states of the prefatigue crack tip by unloading or annealing after the formation of a pre-fatigue crack

Sulfidation/oxidation resistant alloys

International Nuclear Information System (INIS)

Smith, G.D.; Tassen, C.S.

1989-01-01

The patent describes a nickel-base, high chromium alloy. It is characterized by excellent resistance to sulfidation and oxidation at elevated temperatures as high as 2000 degrees F. (1093 degrees C.) and higher, a stress-rupture life of about 200 hours or more at a temperature at least as high as 1800 degrees F. (990:0083 degrees C.) and under a stress of 2000 psi, good tensile strength and good ductility both at room and elevated temperature. The alloy consists essentially of about 27 to 35% chromium, about 2.5 to 5% aluminum, about 2.5 to about 6% iron, 0.5 to 2.5% columbium, up to 0.1% carbon, up to 1% each of titanium and zirconium, up to 0.05% cerium, up to 0.05% yttrium, up to 1% silicon, up to 1% manganese, and the balance nickel

Electrochemical and surface characterization of a nickel-titanium alloy

NARCIS (Netherlands)

Wever, Dirk; Veldhuizen, AG; de Vries, J; Busscher, HJ; Uges, DRA; van Horn, James

1998-01-01

For clinical implantation purposes of shape memory metals the nearly equiatomic nickel-titanium (NiTi) alloy is generally used. In this study, the corrosion properties and surface characteristics of this alloy were investigated and compared with two reference controls, AISI 316 LVM stainless steel

Anelastic relaxation peaks in single crystals of zirconium-oxygen alloys

International Nuclear Information System (INIS)

Ritchie, I.G.; Sprungmann, K.W.; Atrens, A.; Rosinger, H.E.; CEA Centre d'Etudes Nucleaires de Grenoble, 38

1977-01-01

Relaxations of the compliances S 11 -S 12 and S 44 have been observed in single crystals of zirconium-oxygen alloys tested in flexure and in torsion respectively. The relaxations are attributed to the stress-induced reorientation of substitutional impurity atoms (s) paired with interstitial oxygen atoms (i). The results demonstrate that the jump of the interstitial parallel to the basal plane dominates in the reorientation of the s-i pair

Development of new zirconium alloys for PWR fuel rod claddings

International Nuclear Information System (INIS)

Zhao Wenjin; Zhou Bangxin; Miao Zhi; Li Cong; Jiang Hongman; Yu Xiaowei; Jiang Yourong; Huang Qiang; Gou Yuan; Huang Decheng

2001-01-01

An advanced zirconium alloys containing Sn, Nb, Fe and Cr have been developed. The relationships between manufacturing, microstructure and corrosion performance for the new alloys have been studied. The effects of both heat treatment and chemistry on corrosion behavior were assessed by autoclave tests in lithia water at 633 K and high-temperature steam at 773 K. Analytical electron microscopy demonstrated that the best out-of-pile corrosion performance was obtained for microstructure containing a fine and uniform distribution of β-Nb and Zr(Fe, Nb) 2 particles. Autoclave testing in LiOH solution indicated that two kinds of alloys (N18, N36) showed the lower corrosion rate than the reference Zr-4 tested, and especially, the corrosion resistance in superheated steam at 773 K was much better. Moreover, the mechanical properties were superior to Zr-4. And the hydrogen absorption data for all of alloys from corrosion reactions under various corrosion conditions showed a linear increase with the oxide thickness

Status and task of the study on the hydrogen embrittlement of zirconium alloys

International Nuclear Information System (INIS)

Nagase, Fumihisa; Furuta, Teruo; Seino, Shun; Komatsu, Kazushi.

1995-08-01

As the burnup of the LWR fuel is extended, waterside corrosion and hydrogen pickup increase in the Zircaloy cladding. Hydrogen embrittlement of Zircaloy is one of the main factors which may limit the life of the fuel rod. This report presents a review on the hydrogen embrittlement of zirconium and its alloys including the irradiated materials. Research tasks for the reduction of ductility in the high burnup fuel cladding are also discussed. Many fundamental investigations have been performed on the hydrogen embrittlement of zirconium alloys. However, the embrittlement mechanism of the high burnup fuel cladding is complicated. Especially, a coupled effect of hydrides and radiation defects are expected to be pronounced with neutron dose increase. In order to evaluate the reduction of ductility of the higher burnup fuel cladding properly, it is necessary to investigate the coupled effect of these two factors by systematic examinations. (author) 64 refs

Studies on neutron irradiation effects of iron alloys and nickel-base heat resistant alloys

International Nuclear Information System (INIS)

Watanabe, Katsutoshi

1987-09-01

The present paper describes the results of neutron irradiation effects on iron alloys and nickel-base heat resistant alloys. As for the iron alloys, irradiation hardening and embrittlement were investigated using internal friction measurement, electron microscopy and tensile testings. The role of alloying elements was also investigated to understand the irradiation behavior of iron alloys. The essential factors affecting irradiation hardening and embrittlement were thus clarified. On the other hand, postirradiation tensile and creep properties were measured of Hastelloy X alloy. Irradiation behavior at elevated temperatures is discussed. (author)

Mechanodynamical analysis of nickel-titanium alloys for orthodontics application

International Nuclear Information System (INIS)

Arruda, Carlos do Canto

2002-01-01

Nickel-titanium alloys may coexist in more than one crystalline structure. There is a high temperature phase, austenite, and a low temperature phase, martensite. The metallurgical basis for the superelasticity and the shape memory effect relies in the ability of these alloys to transform easily from one phase to another. There are three essential factors for the orthodontist to understand nickel-titanium alloys behaviour: stress; deflection; and temperature. These three factors are related to each other by the stress-deflection, stress-temperature and deflection-temperature diagrams. This work was undertaken with the objective to analyse commercial nickel-titanium alloys for orthodontics application, using the dynamical mechanical analyser - DMA. Four NiTi 0,017 X 0,025'' archwires were studied. The archwires were Copper NiTi 35 deg C (Ormco), Neo Sentalloy F200 (GAC), Nitinol Superelastic (Unitek) and NiTi (GAC). The different mechanodynamical properties such as elasticity and damping moduli were evaluated. Each commercial material was evaluated with and without a 1 N static force, aiming to evaluate phase transition temperature variation with stress. The austenitic to martensitic phase ratio, for the experiments without static force, was in the range of 1.59 to 1.85. For the 1 N static force tests the austenitic to martensitic phase ratio, ranged from 1.28 to 1.57 due to the higher martensite elasticity modulus. With elastic modulus variation with temperature behaviour, the orthodontist has the knowledge of the force variation applied in the tooth in relation to the oral cavity temperature change, for nickel-titanium alloys that undergo phase transformation. The damping capacity of the studied alloys depends on the materials state: martensitic phase; austenitic phase or during phase transformation. The martensitic phase shows higher dumping capacity. During phase transformation, an internal friction peak may be observed for the CuNiTi 35 deg C and Neo Sentalloy F

Activation analysis in zirconium and alloys for nuclear application

International Nuclear Information System (INIS)

Cohen, I.M.; Mila, M.I.; Gomez, C.D.

1981-01-01

A study has been performed with the purpose to ascertain the possibilities of using neutron activation analysis in non-destructive determination of several elements present in zirconium and its alloys. Those elements must be limited within acceptable top levels, in accordance to standards for nuclear applications. The experimental techniques used are described and the results obtained are discussed, showing that the method is adequate for determining Cl, Co, Hf, Mn, and W, but not Ni and U. (M.E.L.) [es

Effect of Exogenous Zirconia Nanophases on the Structural Properties of the Sulfur- and Tin-Containing Nickel Melts

Science.gov (United States)

Anuchkin, S. N.

2017-11-01

The surface tension and the density of the nickel melts with introduced ZrO2 nanoparticles are studied by the sessile drop method using a digital camera and computer processing of images. The revealed differently directed effects of nanoparticles on the surface tension in the Ni-Sn and Ni-S systems points to a change in the structure of the melt-gas surface layer. The nanoparticles are shown to affect the adsorption of surfactants, and the surface layer is likely to consist of adsorbed Ni + (ZrO2-surfactant) ensembles. The ZrO2 content in a metal is determined using the technique of separate determination of the zirconium content dissolved in a metal and zirconium in the form of ZrO2. It was found that, at 0.10 wt % ZrO2 initially present in a metal, 0.021-0.031 wt % ZrO2 are retained in samples; that is, about 70 rel % ZrO2 are removed to the interface in the form of ensembles. Auger spectroscopy analysis of the Ni-Sn-ZrO2 surface film detected 5-10 rel % Zr in the surface layer.

Austenitic alloys Fe-Ni-Cr dominating

International Nuclear Information System (INIS)

Gibson, R.C.; Korenko, M.K.

1980-01-01

Austenitic alloy essentially comprising 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% aluminium, 0.1 to 0.3% titanium, 0.02 to 0.05% carbon, 0.002 to 0.015% boron, up to 0.06% zirconium, the balance being iron. The characteristic of this alloy is a conventional elasticity limit to within 2% of at least 450 MPa, with a maximum tensile strength of at least 500 MPa at a test temperature of 650 0 C after immersion annealing at 1038 0 C and 30% hardening. To this effect the invention concerns Ni-Cr-Fe high temperature alloys possessing excellent mechanical strength characteristics, that can be obtained with lower levels of nickel and chromium than those used in alloys of this kind in the present state of the technique, a higher amount of niobium than in the previous alloys and with the addition of 0.5 to 1.5% vanadium [fr

High resisting alloy without Co used in nuclear industry

International Nuclear Information System (INIS)

Balleret, Alain.

1976-01-01

The description is given of a high resistance alloy characterised in that it includes by weight 5 to 14% molybdenum, 19 to 32% chromium, 2 to 8% tungsten, 6 to 50% nickel, 0.2 to 2.8% carbon, 0 to 5% vanadium, 0 to 5% zirconium, 0 to 5% niobium-tantalum, 0 to 3% manganese, 0 to 3% silicon, 0 to 1.5% boron and iron in an amount to ensure the global balance of this alloy [fr

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

Science.gov (United States)

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.

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

International Nuclear Information System (INIS)

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 deg. 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 deg. 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 deg. 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

Corrosion resistance of metals and alloys in molten alkalies

International Nuclear Information System (INIS)

Zarubitskij, O.G.; Dmitruk, B.F.; Minets, L.A.

1979-01-01

Literature data on the corrosion of non-ferrous and noble metals, iron and steels in the molten alkalis and mixtures of their base are presented. It is shown that zirconium, niobium and tantalum are characterized by high corrosion stability in the molten NaOH. Additions of NaOH and KOH to the alkali chloride melts result in a 1000 time decrease of zirconium corrosion rate at 850 deg. The data testify to the characteristic passivating properties of OH - ions; Mo and W do not possess an ability to selfpassivation in hydroxide melts. Corrosion resistance of carbon and chromium-nickel steels in hydroxide melts depends considerably on the temperature, electrolyte composition and atmosphere over them. At the temperatures up to 600 deg C chromium-nickel steel is corrosion resistant in the molten alkali only in the inert atmosphere. Corrosion rate of chromium-nickel alloy is the lower the less chromium and the more nickel it contains. For the small installations the 4Kh18N25S2 and Kh23N28M3D3T steels can be recommended

Characteristics of Pilger Die Materials for Nuclear Zirconium Alloy Tubes

International Nuclear Information System (INIS)

Park, Ki Bum; Kim, In Kyu; Park, Min Young; Kahng, Jong Yeol; Kim, Sun Doo

2011-01-01

KEPCO Nuclear Fuel Company's (KEPCO NF) tube manufacturing facility, Techno Special Alloy (TSA) Plant, has started cold pilgering operation since 2008. It is obvious that the cold pilgering process is one of the key processes controlling the quality and the characteristics of the tubes manufactured, i.e. nuclear zirconium alloy tube in KEPCO NF. Cold pilgering is a rolling process for forming metal tubes in which diameter and wall thickness are reduced in a number of forming steps, using ring dies at outside of the tube and a curved mandrel at inside to reduce tube cross sections by up to 90 percent. The OD size of tube is reduced by a pair of dies, and ID size and wall thickness is controlled simultaneously by mandrel. During the cold pilgering process, both tools are the critical components for providing qualified tube. Development of pilger die and mandrel has been a significant importance in the zirconium tube manufacturing and a major goal of KEPCO NF. The objective of this study is to evaluate the life time of pilger die during pilgering. Therefore, a comparison of the heat treatment and mechanical properties of between AISI 52100 and AISI H13 materials was made in this study

Corrosion resistance of nickel alloys with chromium and silicon to the red fuming nitric acid

International Nuclear Information System (INIS)

Gurvich, L.Ya.; Zhirnov, A.D.

1994-01-01

Corrosion and electrochemical behaviour of binary Ni-Cr, Ni-Si nickel and ternary Ni-Cr-Si alloys in the red fuming nitric acid (RFNA) (8-% of HNO 3 +20% of N 2 O 4 ) is studied. It is shown that nickel alloying with chromium improves its corrosion resistance to the red fuming nitric acid. Nickel alloying with silicon in quantities of up to 5 % reduces, and up to 10%-increases abruptly the corrosion resistance with subsequent decrease of the latter after the further increase of concentration. Ni-15% of Cr alloy alloying with silicon increases monotonously the corrosion resistance. 10 refs., 7 figs., 3 tabs

Hydrogen charging, hydrogen content analysis and metallographic examination of hydride in zirconium alloys

International Nuclear Information System (INIS)

Singh, R.N.; Kishore, R.; Mukherjee, S.; Roychowdhury, S.; Srivastava, D.; Sinha, T.K.; De, P.K.; Banerjee, S.; Gopalan, B.; Kameswaran, R.; Sheelvantra, Smita S.

2003-12-01

Gaseous and electrolytic hydrogen charging techniques for introducing controlled amount of hydrogen in zirconium alloy is described. Zr-1wt%Nb fuel tube, zircaloy-2 pressure tube and Zr-2.5Nb pressure tube samples were charged with up to 1000 ppm of hydrogen by weight using one of the aforementioned methods. These hydrogen charged Zr-alloy samples were analyzed for estimating the total hydrogen content using inert gas fusion technique. Influence of sample surface preparation on the estimated hydrogen content is also discussed. In zirconium alloys, hydrogen in excess of the terminal solid solubility precipitates out as brittle hydride phase, which acquire platelet shaped morphology due to its accommodation in the matrix and can make the host matrix brittle. The F N number, which represents susceptibility of Zr-alloy tubes to hydride embrittlement was measured from the metallographs. The volume fraction of the hydride phase, platelet size, distribution, interplatelet spacing and orientation were examined metallographically using samples sliced along the radial-axial and radial-circumferential plane of the tubes. It was observed that hydride platelet length increases with increase in hydrogen content. Considering the metallographs generated by Materials Science Division as standard, metallographs prepared by the IAEA round robin participants for different hydrogen concentration was compared. It is felt that hydride micrographs can be used to estimate not only that approximate hydrogen concentration of the sample but also its size, distribution and orientation which significantly affect the susceptibility to hydride embrittlement of these alloys. (author)

Microstructure and corrosion behavior of electrodeposited nano-crystalline nickel coating on AZ91 Mg alloy

Energy Technology Data Exchange (ETDEWEB)

Zarebidaki, Arman, E-mail: arman.zare@iauyazd.ac.ir; Mahmoudikohani, Hassan, E-mail: hassanmahmoudi.k@gmail.com; Aboutalebi, Mohammad-Reza

2014-12-05

Highlights: • Activation, zincating, and Cu electrodeposition were used as pretreatment processes for electrodeposition of nickel coatings. • Nano-crystalline nickel coatings were successfully electrodeposited onto the AZ91 Mg alloys. • Effect of nickel electrodeposited coating on the corrosion resistance of AZ91 Mg alloy has been studied. - Abstract: In order to enhance the corrosion resistance, nickel coating was electrodeposited onto AZ91 Mg alloy. Activation, zincating, and Cu electrodeposition used as pretreatment processes for better adhesion and corrosion performance of the nickel over layer. The corrosion properties of the AZ91 Mg alloy, nickel electroplated AZ91 Mg alloy, and pure nickel was assessed via polarization and electrochemical impedance spectroscopy (EIS) methods in 3.5 wt% NaCl solution. Moreover, the structure of the coating was investigated by means of X-ray diffraction, whereas specimen’s morphology and elemental composition were analyzed using scanning electron microscope (SEM) equipped with energy dispersive spectrometer (EDS). Measurements revealed that the coating has a nano-crystalline structure with the grain size of 95 nm. Corrosion results showed superior corrosion resistance for the coated AZ91 Mg alloy as the corrosion current density decreased from 2.5 × 10{sup −4} A cm{sup −2}, for the uncoated sample, to 1.5 × 10{sup −5} A cm{sup −2}, for coated specimen and the corrosion potential increased from −1.55 V to −0.98 V (vs. Ag/AgCl) at the same condition.

Process for electrolytic deposition of metals on zirconium materials

International Nuclear Information System (INIS)

Donaghy, R.E.

1981-01-01

An article made of a zirconium alloy can be electrolytically plated with a layer of a metal such as copper, nickel or chromium when the article is free of any loosely adhering film formed during an activation step. The article is activated in an aged aqueous solution of ammonium bifluoride and sulfuric acid. Next the loosely adhering film formed in the first step is removed by chemical treatment, ultrasonic cleaning, or by swabbing the surface with cotton or an organic material. Finally the article is contacted with an electrolytic plating solution in the presence of an electrode receiving current

Polycrystalline models for the calculation of residual stresses in zirconium alloys tubes

International Nuclear Information System (INIS)

Signorelli, J.W.; Turner, P.A.; Lebensohn, R.A.; Pochettino, A.A.

1995-01-01

Tubes made of different Zirconium alloys are used in various types of reactors. The final texture of tubes as well as the distribution of residual stresses depend on the mechanical treatments done during their manufacturing process. The knowledge and prediction of both the final texture and the distribution of residual stresses in a tube for nuclear applications are of outstanding importance in relation with in-reactor performance of the tube, especially in what concerns to its irradiation creep and growth behaviour. The viscoplastic and the elastoplastic self consistent polycrystal models are used to investigate the influence of different mechanical treatments, performed during rolling processes on the final distribution of intergranular residual stresses of zirconium alloys tubes. The residual strains predictions with both formulations show a non linear dependence with the orientation, but they are qualitatively different. This discrepancy could be explain in terms of the relative plastic activity between the -type and -type deformation modes predicted with the viscoplastic and elastoplastic models. (author). 10 refs., 4 figs., 1 tab

Graph theory and binary alloys passivated by nickel

International Nuclear Information System (INIS)

McCafferty, E.

2005-01-01

The passivity of a nickel binary alloy is considered in terms of a network of -Ni-O-Ni- bridges in the oxide film, where Ni is the component of the binary alloy which produces passivity. The structure of the oxide is represented by a mathematical graph, and graph theory is used to calculate the connectivity of the oxide, given by the product of the number of edges in the graph and the Randic index. A stochastic calculation is employed to insert ions of the second metal into the oxide film so as to disrupt the connectivity of the -Ni-O-Ni- network. This disruption occurs at a critical ionic concentration of the oxide film. Mathematical relationships are developed for the introduction of a general ion B +n into the oxide film, and critical ionic compositions are calculated for oxide films on the nickel binary alloys. The notation B refers to any metal B which produces B +n ions in the oxide film, where +n is the oxidation number of the ion. The results of this analysis for Fe-Ni and Cu-Ni binary alloys are in good agreement with experimental results

Development of high-capacity nickel-metal hydride batteries using superlattice hydrogen-absorbing alloys

International Nuclear Information System (INIS)

Yasuoka, Shigekazu; Magari, Yoshifumi; Murata, Tetsuyuki; Tanaka, Tadayoshi; Ishida, Jun; Nakamura, Hiroshi; Nohma, Toshiyuki; Kihara, Masaru; Baba, Yoshitaka; Teraoka, Hirohito

2006-01-01

New R-Mg-Ni (R: rare earths) superlattice alloys with higher-capacity and higher-durability than the conventional Mm-Ni alloys with CaCu 5 structure have been developed. The oxidation resistibility of the superlattice alloys has been improved by optimizing the alloy composition by such as substituting aluminum for nickel and optimizing the magnesium content in order to prolong the battery life. High-capacity nickel-metal hydride batteries for the retail market, the Ni-MH2500/900 series (AA size type 2500mAh, AAA size type 900mAh), have been developed and commercialized by using an improved superlattice alloy for negative electrode material. alized by using an improved superlattice alloy for negative electrode material. (author)

Influence of Chromium and Molybdenum on the Corrosion of Nickel Based Alloys

International Nuclear Information System (INIS)

Hayes, J R; Gray, J; Szmodis, A W; Orme, C A

2005-01-01

The addition of chromium and molybdenum to nickel creates alloys with exceptional corrosion resistance in a diverse range of environments. This study examines the complementary roles of Cr and Mo in Ni alloy passivation. Four nickel alloys with varying amounts of chromium and molybdenum were studied in 1 molar salt solutions over a broad pH range. The passive corrosion and breakdown behavior of the alloys suggests that chromium is the primary element influencing general corrosion resistance. The breakdown potential was nearly independent of molybdenum content, while the repassivation potential is strongly dependant on the molybdenum content. This indicates that chromium plays a strong role in maintaining the passivity of the alloy, while molybdenum acts to stabilize the passive film after a localized breakdown event

Development of high-capacity nickel-metal hydride batteries using superlattice hydrogen-absorbing alloys

Science.gov (United States)

Yasuoka, Shigekazu; Magari, Yoshifumi; Murata, Tetsuyuki; Tanaka, Tadayoshi; Ishida, Jun; Nakamura, Hiroshi; Nohma, Toshiyuki; Kihara, Masaru; Baba, Yoshitaka; Teraoka, Hirohito

New R-Mg-Ni (R: rare earths) superlattice alloys with higher-capacity and higher-durability than the conventional Mm-Ni alloys with CaCu 5 structure have been developed. The oxidation resistibility of the superlattice alloys has been improved by optimizing the alloy composition by such as substituting aluminum for nickel and optimizing the magnesium content in order to prolong the battery life. High-capacity nickel-metal hydride batteries for the retail market, the Ni-MH2500/900 series (AA size type 2500 mAh, AAA size type 900 mAh), have been developed and commercialized by using an improved superlattice alloy for negative electrode material.

Effects of titanium and zirconium on iron aluminide weldments

Energy Technology Data Exchange (ETDEWEB)

Mulac, B.L.; Edwards, G.R. [Colorado School of Mines, Golden, CO (United States). Center for Welding, Joining, and Coatings Research; Burt, R.P. [Alumax Technical Center, Golden, CO (United States); David, S.A. [Oak Ridge National Lab., TN (United States). Metals and Ceramics Div.

1997-12-01

When gas-tungsten arc welded, iron aluminides form a coarse fusion zone microstructure which is susceptible to hydrogen embrittlement. Titanium inoculation effectively refined the fusion zone microstructure in iron aluminide weldments, but the inoculated weldments had a reduced fracture strength despite the presence of a finer microstructure. The weldments fractured by transgranular cleavage which nucleated at cracked second phase particles. With titanium inoculation, second phase particles in the fusion zone changed shape and also became more concentrated at the grain boundaries, which increased the particle spacing in the fusion zone. The observed decrease in fracture strength with titanium inoculation was attributed to increased spacing of second phase particles in the fusion zone. Current research has focused on the weldability of zirconium- and carbon-alloyed iron aluminides. Preliminary work performed at Oak Ridge National Laboratory has shown that zirconium and carbon additions affect the weldability of the alloy as well as the mechanical properties and fracture behavior of the weldments. A sigmajig hot cracking test apparatus has been constructed and tested at Colorado School of Mines. Preliminary characterization of hot cracking of three zirconium- and carbon-alloyed iron aluminides, each containing a different total concentration of zirconium at a constant zirconium/carbon ratio of ten, is in progress. Future testing will include low zirconium alloys at zirconium/carbon ratios of five and one, as well as high zirconium alloys (1.5 to 2.0 atomic percent) at zirconium/carbon ratios of ten to forty.

Magnetic properties of the binary Nickel/Bismuth alloy

Energy Technology Data Exchange (ETDEWEB)

Keskin, Mustafa; Şarlı, Numan, E-mail: numansarli82@gmail.com

2017-09-01

Highlights: • We model and investigate the magnetic properties of the Ni/Bi alloy within the EFT. • Magnetizations of the Ni/Bi alloy are observed as Bi1 > Bi2 > Ni/Bi > Ni at T < Tc. • Magnetization of the Bi1 is dominant and Ni is at least dominant T < Tc. • Total magnetization of the Ni/Bi alloy is close to those of Ni at T < Tc. • Hysteresis curves are overlap at T < 0.1 and they behave separately at T > 0.1. - Abstract: Magnetic properties of the binary Nickel/Bismuth alloy (Ni/Bi) are investigated within the effective field theory. The Ni/Bi alloy has been modeled that the rhombohedral Bi lattice is surrounded by the hexagonal Ni lattice. According to lattice locations, Bi atoms have two different magnetic properties. Bi1 atoms are in the center of the hexagonal Ni atoms (Ni/Bi1 single layer) and Bi2 atoms are between two Ni/Bi1 bilayers. The Ni, Bi1, Bi2 and Ni/Bi undergo a second-order phase transition from the ferromagnetic phase to paramagnetic phase at Tc = 1.14. The magnetizations of the Ni/Bi alloy are observed as Bi1 > Bi2 > Ni/Bi > Ni at T < Tc; hence the magnetization of the Bi1 is dominant and Ni is at least dominant. However, the total magnetization of the Ni/Bi alloy is close to magnetization of the Ni at T < Tc. The corcivities of the Ni, Bi1, Bi2 and Ni/Bi alloy are the same with each others, but the remanence magnetizations are different. Our theoretical results of M(T) and M(H) of the Ni/Bi alloy are in quantitatively good agreement with the some experimental results of binary Nickel/Bismuth systems.

Stainless steel-zirconium alloy waste forms

International Nuclear Information System (INIS)

McDeavitt, S.M.; Abraham, D.P.; Keiser, D.D. Jr.; Park, J.Y.

1996-01-01

An electrometallurgical treatment process has been developed by Argonne National Laboratory to convert various types of spent nuclear fuels into stable storage forms and waste forms for repository disposal. The first application of this process will be to treat spent fuel alloys from the Experimental Breeder Reactor-II. Three distinct product streams emanate from the electrorefining process: (1) refined uranium; (2) fission products and actinides extracted from the electrolyte salt that are processed into a mineral waste form; and (3) metallic wastes left behind at the completion of the electrorefining step. The third product stream (i.e., the metal waste stream) is the subject of this paper. The metal waste stream contains components of the chopped spent fuel that are unaffected by the electrorefining process because of their electrochemically ''noble'' nature; this includes the cladding hulls, noble metal fission products (NMFP), and, in specific cases, zirconium from metal fuel alloys. The selected method for the consolidation and stabilization of the metal waste stream is melting and casting into a uniform, corrosion-resistant alloy. The waste form casting process will be carried out in a controlled-atmosphere furnace at high temperatures with a molten salt flux. Spent fuels with both stainless steel and Zircaloy cladding are being evaluated for treatment; thus, stainless steel-rich and Zircaloy-rich waste forms are being developed. Although the primary disposition option for the actinides is the mineral waste form, the concept of incorporating the TRU-bearing product into the metal waste form has enough potential to warrant investigation

In-situ electrochemical impedance spectroscopy measurements of zirconium alloy oxide conductivity: Relationship to hydrogen pickup

International Nuclear Information System (INIS)

Couet, Adrien; Motta, Arthur T.; Ambard, Antoine; Livigni, Didier

2017-01-01

Highlights: • In-situ electrochemistry on zirconium alloys in 360 °C pure water show oxide layer resistivity changes during corrosion. • A linear relationship is observed between oxide resistivity and instantaneous hydrogen pickup fraction. • The resistivity of the oxide layer formed on Zircaloy-4 (and thus its hydrogen pickup fraction) is higher than on Zr-2.5Nb. - Abstract: Hydrogen pickup during nuclear fuel cladding corrosion is a critical life-limiting degradation mechanism for nuclear fuel. Following a program dedicated to zirconium alloys, corrosion, it has been hypothesized that oxide electronic resistivity determines hydrogen pickup. In-situ electrochemical impedance spectroscopy experiments were performed on Zircaloy-4 and Zr-2.5Nb alloys in 360 °C water. The oxide resistivity was measured as function of time. The results show that as the oxide resistivity increases so does the hydrogen pickup fraction. The resistivity of the oxide layer formed on Zircaloy-4 is higher than on Zr-2.5Nb, resulting in a higher hydrogen pickup fraction of Zircaloy-4, compared to Zr-2.5Nb.

CASTI handbook of stainless steels and nickel alloys. 2. ed.

International Nuclear Information System (INIS)

Lamb, S.

2002-01-01

This is the only up-to-date (2002) reference book that covers both stainless steels and nickel alloys. Written by 30 authors and peer reviewers with over 700 years of combined industrial experience, this CASTI handbook provides the latest stainless steels and nickel alloys information in a practical and comprehensive manner. For the project engineer, maintenance engineer or inspector, this book provides solutions to many of the corrosion problems encountered in aggressive environmental conditions. Some of the corrosive conditions covered are: stress corrosion cracking, reducing environments, halogenation, highly oxidizing environments, and high temperatures. Hundreds of different material applications and selections, throughout many industries, are referenced. It is an ideal reference source to assist in preventing or minimizing corrosion related problems, including those encountered during welding fabrication. This practical handbook also contains a handy 'Alloy Index' which lists each alloy by its ASTM Specification, UNS Number, common name, trade name and page number references. The second edition includes additional coverage of corrosion resistant alloys for downhole production tubing. The new material covers corrosion processes, corrosion rates, hydrogen sulfide environments, corrosion inhibitors, corrosion resistant alloys, the application of stainless steel in production conditions, and more

Thermal cycling influence on microstructural characterization of alloys with high nickel content

International Nuclear Information System (INIS)

Abrudeanu, M.; Gradin, O.; Vulpe, S. C.; Ohai, D.

2013-01-01

The IV nuclear energy generation systems are aimed at making revolutionary improvements in economics, safety and reliability, and sustainability. To achieve these goals, Generation IV systems will operate at higher temperatures and in higher radiation fields. This paper shows the thermal cycling influences on microstructure and hardness of nickel based alloys: Incoloy 800 HT and Inconel 617. These alloys were meekly at a thermal cycling of 25, 50, 75 and 100 cycles. The temperature range of a cycle was between 400 O C and 700 O C. Nickel base alloys develop their properties by solid solution and/or precipitation strengthening. (authors)

Generation of copper, nickel, and CuNi alloy nanoparticles by spark discharge

International Nuclear Information System (INIS)

Muntean, Alex; Wagner, Moritz; Meyer, Jörg; Seipenbusch, Martin

2016-01-01

The generation of copper, nickel, and copper-nickel alloy nanoparticles by spark discharge was studied, using different bespoke alloy feedstocks. Roughly spherical particles with a primary particle Feret diameter of 2–10 nm were produced and collected in agglomerate form. The copper-to-nickel ratios determined by Inductively coupled plasma mass spectrometry (ICP-MS), and therefore averaged over a large number of particles, matched the nominal copper content quite well. Further investigations showed that the electrode compositions influenced the evaporation rate and the primary particle size. The evaporation rate decreased with increasing copper content, which was found to be in good accordance with the Llewellyn-Jones model. However, the particle diameter was increasing with an increasing copper content, caused by a decrease in melting temperature due to the lower melting point of copper. Furthermore, the alloy compositions on the nanoscale were investigated via EDX. The nanoparticles exhibited almost the same composition as the used alloy feedstock, with a deviation of less than 7 percentage points. Therefore, no segregation could be detected, indicating the presence of a true alloy even on the nanoscale.

Acoustic emission from zirconium alloys during mechanical and fracture testing

International Nuclear Information System (INIS)

Coleman, C.E.

1986-10-01

The application of acoustic emission during the mechanical and fracture testing of zirconium alloys is reviewed. Acoustic emission is successful in following delayed hydride cracking quantitatively. It is especially useful when great sensitivity is required. Application to fatigue, tensile deformation and stress corrosion cracking appears promising but requires more work to separate phenomena before it can be used quantitatively. This report is based on an invited review for the American Society of Non-Destructive Testing Handbook: Volume 5, Acoustic Emission Testing

Experimental and thermodynamic study of the erbium-oxygen-zirconium and gadolinium-oxygen-zirconium systems

International Nuclear Information System (INIS)

Jourdan, J.

2009-11-01

This work is a contribution to the development of innovative concepts for fuel cladding in pressurized water nuclear reactors. This concept implies the insertion of rare earth (erbium and gadolinium) in the zirconium fuel cladding. The determination of phase equilibria in the systems is essential prior to the implementation of such a promising solution. This study consisted in an experimental determination of the erbium-zirconium phase diagram. For this, we used many different techniques in order to obtain diagram data such as solubility limits, solidus, liquidus or invariant temperatures. These data allowed us to present a new diagram, very different from the previous one available in the literature. We also assessed the diagram using the CALPHAD approach. In the gadolinium-zirconium system, we determined experimentally the solubility limits. Those limits had never been determined before, and the values we obtained showed a very good agreement with the experimental and assessed versions of the diagram. Because these alloys are subjected to oxygen diffusion throughout their life, we focused our attention on the erbium-oxygen-zirconium and gadolinium-oxygen-zirconium systems. The first system has been investigated experimentally. The alloys fabrication has been performed using powder metallurgy. In order to obtain pure raw materials, we fabricated powder from erbium and zirconium bulk metals using hydrogen absorption/desorption. The characterisation of the ternary pellets allowed the determination of two ternary isothermal sections at 800 and 1100 C. For the gadolinium-oxygen-zirconium system, we calculated the phase equilibria at temperatures ranging from 800 to 1100 C, using a homemade database compiled from literature assessments of the oxygen-zirconium, gadolinium-zirconium and gadolinia-zirconia systems. Finally, we determined the mechanical properties, in connexion with the microstructure, of industrial quality alloys in order to identify the influence of

Electrodeposition of zinc-nickel alloy from fluoborate baths - as a substitute for electrogalvanising

Energy Technology Data Exchange (ETDEWEB)

Ramesh Bapu, G.N.K.; Ayyapparaju, J.; Devaraj, G.

Use of fluoborate electroytes have been investigated for depositing a suitable composition of zinc-nickel alloy on mild steel for better corrosion protection. In the present investigation, the plating and bath conditions have been optimized so that zinc-nickel alloy coating from fluoborate solutions find applications for plating wires as well as other articles advantageously in the place of zinc coatings.

Corrosion of AZ91D magnesium alloy with a chemical conversion coating and electroless nickel layer

International Nuclear Information System (INIS)

Huo Hongwei; Li Ying; Wang Fuhui

2004-01-01

A chemical conversion treatment and an electroless nickel plating were applied to AZ91D alloy to improve its corrosion resistance. By conversion treatment in alkaline stannate solution, the corrosion resistance of the alloy was improved to some extent as verified by immersion test and potentiodynamic polarization test in 3.5 wt.% NaCl solution at pH 7.0. X-ray diffraction patterns of the stannate treated AZ91D alloy showed the presence of MgSnO 3 · H 2 O, and SEM images indicated a porous structure, which provided advantage for the adsorption during sensitisation treatment prior to electroless nickel plating. A nickel coating with high phosphorus content was successfully deposited on the chemical conversion coating pre-applied to AZ91D alloy. The presence of the conversion coating between the nickel coating and the substrate reduced the potential difference between them and enhanced the corrosion resistance of the alloy. An obvious passivation occurred for the nickel coating during anodic polarization in 3.5 wt.% NaCl solution

Standard specification for Nickel-Chromium-Iron alloys (UNS N06600, N06601, N06603, N06690, N06693, N06025, N06045, and N06696), Nikel-Chromium-Cobalt-Molybdenum alloy (UNS N06617), and Nickel-Iron-Chromium-Tungsten alloy (UNS N06674) seamless pipe and tube

CERN Document Server

American Society for Testing and Materials. Philadelphia

2011-01-01

Standard specification for Nickel-Chromium-Iron alloys (UNS N06600, N06601, N06603, N06690, N06693, N06025, N06045, and N06696), Nikel-Chromium-Cobalt-Molybdenum alloy (UNS N06617), and Nickel-Iron-Chromium-Tungsten alloy (UNS N06674) seamless pipe and tube

Phases in lanthanum-nickel-aluminum alloys

International Nuclear Information System (INIS)

Mosley, W.C.

1992-01-01

Lanthanum-nickel-aluminum (LANA) alloys will be used to pump, store and separate hydrogen isotopes in the Replacement Tritium Facility (RTF). The aluminum content (y) of the primary LaNi 5 -phase is controlled to produce the desired pressure-temperature behavior for adsorption and desorption of hydrogen. However, secondary phases cause decreased capacity and some may cause undesirable retention of tritium. Twenty-three alloys purchased from Ergenics, Inc. for development of RTF processes have been characterized by scanning electron microscopy (SEM) and by electron microprobe analysis (EMPA) to determine the distributions and compositions of constituent phases. This memorandum reports the results of these characterization studies. Knowledge of the structural characteristics of these alloys is a useful first step in selecting materials for specific process development tests and in interpreting results of those tests. Once this information is coupled with data on hydrogen plateau pressures, retention and capacity, secondary phase limits for RTF alloys can be specified

Methods of studying oxide scales grown on zirconium alloys in autoclaves and in a PWR

International Nuclear Information System (INIS)

Blank, H.; Bart, G.; Thiele, H.

1992-01-01

The analysis of water-side corrosion of zirconium alloys has been a field of research for more than 25 years, but the details of the mechanisms involved still cannot be put into a coherent picture. Improved methods are required to establish the details of the microstructure of the oxide scales. A new approach has been made for a general analysis of oxide specimens from scales grown on the zirconium-based cladding alloys of PWR rods in order to analyse the morphology of these scales, the topography of the oxide/metal interface and the crystal structures close to this interface: a) Instead of using the conventional pickling solutions, the Zr-alloys are dissolved using a 'softer' solution (Br 2 in an organic solvent) in order to avoid damage to the oxide at the oxide/metal interface to be analysed by SEM (scanning electron microscopy). A second advantage of this method is easy etching of the grain structure of Zr-alloys for SEM analysis; b) By using the particular properties of the oxide scales, the corrosion-rate-determining innermost part of the oxide layer at the oxide/metal interface can be separated from the rest of the oxide scale and then analysed by SEM, STEM (scanning transmission electron microscopy), TEM (transmission electron microscopy) and electron diffraction after dissolution of the alloy. Examples are given from oxides grown on Zr-alloys in a pressurized water reactor and in autoclaves. (author) 8 figs., 3 tabs., 9 refs

Nickel-titanium alloys: stress-related temperature transitional range.

Science.gov (United States)

Santoro, M; Beshers, D N

2000-12-01

The inducement of mechanical stress within nickel-titanium wires can influence the transitional temperature range of the alloy and therefore the expression of the superelastic properties. An analogous variation of the transitional temperature range may be expected during orthodontic therapy, when the archwires are engaged into the brackets. To investigate this possibility, samples of currently used orthodontic nickel-titanium wires (Sentalloy, GAC; Copper Ni-Ti superelastic at 27 degrees C, 35 degrees C, 40 degrees C, Ormco; Nitinol Heat-Activated, 3M-Unitek) were subjected to temperature cycles ranging between 4 degrees C and 60 degrees C. The wires were mounted in a plexiglass loading device designed to simulate clinical situations of minimum and severe dental crowding. Electrical resistivity was used to monitor the phase transformations. The data were analyzed with paired t tests. The results confirmed the presence of displacements of the transitional temperature ranges toward higher temperatures when stress was induced. Because nickel-titanium wires are most commonly used during the aligning stage in cases of severe dental crowding, particular attention was given to the performance of the orthodontic wires under maximum loading. An alloy with a stress-related transitional temperature range corresponding to the fluctuations of the oral temperature should express superelastic properties more consistently than others. According to our results, Copper Ni-Ti 27 degrees C and Nitinol Heat-Activated wires may be considered suitable alloys for the alignment stage.

Voltammetric determination of zirconium using azo compounds

International Nuclear Information System (INIS)

Orshulyak, O.O.; Levitskaya, G.D.

2008-01-01

The optimum conditions for zirconium complexation with azo compounds are found. The applicability of Eriochrome Red B, Calcon, and Calcion to the voltammetric determination of zirconium, total Zr(IV) and Hf(IV), and Zr(IV) in the presence of Zn(II), Cu(II), Cd(II), Ni(II), or Ti(IV) is demonstrated. The developed procedures are used to determine zirconium in a terbium alloy and in an alloy for airplane wheel drums [ru

Irradiation growth in zirconium alloys: a review

International Nuclear Information System (INIS)

Fidleris, V.

1980-09-01

The change in shape during irradiation without external stress, irradiation growth, was first discovered in uranium and later in graphite, zirconium and other core materials which exhibit anisotropic physical properties. The direction of maximum growth of metals invariably corresponds with the direction of minimum thermal expansion. In polycrystalline zirconium alloys growth is positive in the direction of maximum deformation during fabrication and in other directions it can be either positive or negative depending on the preferred orientation of grains (crystallographic texture). Growth increases gradually with temperature between 300 K and 620 K and rapidly with fluence up to about 1 x 10 25 n.m. -2 (Eμ1 MeV). At higher fluences the growth appears to saturate in annealed materials and reach a steady rate approximately proportional to dislocation density in cold-worked materials. Above 600 K both annealed and cold-worked materials have similar steady growth rates. Irradiation growth is caused by the segregation to different sinks of the vacancies and interstitials generated by irradiation, but the dominant types of sinks for each type of point defect and the mode of transport of the point defects to sinks cannot therefore be predicted theoretically. For the purpose of designing reactor core components empirical equations have been derived that can satisfactorily predict the steady state growth behaviour from texture and microstructure. (auth)

Ultrasonic texture characterization of aluminum, zirconium and titanium alloys

International Nuclear Information System (INIS)

Anderson, A.J.

1997-01-01

This work attempts to show the feasibility of nondestructive characterization of non-ferrous alloys. Aluminum alloys have a small single crystal anisotropy which requires very precise ultrasonic velocity measurements for derivation of orientation distribution coefficients (ODCs); the precision in the ultrasonic velocity measurement required for aluminum alloys is much greater than is necessary for iron alloys or other alloys with a large single crystal anisotropy. To provide greater precision, some signal processing corrections need to be applied to account for the inherent, half-bandwidth offset in triggered pulses when using a zero-crossing technique for determining ultrasonic velocity. In addition, alloys with small single crystal anisotropy show a larger dependence on the single crystal elastic constants (SCECs) when predicting ODCs which require absolute velocity measurements. Attempts were made to independently determine these elastics constants in an effort to improve correlation between ultrasonically derived ODCs and diffraction derived ODCs. The greater precision required to accurately derive ODCs in aluminum alloys using ultrasonic nondestructive techniques is easily attainable. Ultrasonically derived ODCs show good correlation with derivations made by Bragg diffraction techniques, both neutron and X-ray. The best correlation was shown when relative velocity measurements could be used in the derivations of the ODCs. Calculation of ODCs in materials with hexagonal crystallites can also be done. Because of the crystallite symmetries, more information can be extracted using ultrasonic techniques, but at a cost of requiring more physical measurements. Some industries which use materials with hexagonal crystallites, e.g. zirconium alloys and titanium, have traditionally used texture parameters which provide some specialized measure of the texture. These texture parameters, called Kearns factors, can be directly related to ODCs

Effect of high hydrogen content on metallurgical and mechanical properties of zirconium alloy claddings after heat-treatment at high temperature

International Nuclear Information System (INIS)

Turque, Isabelle

2016-01-01

Under hypothetical loss-of-coolant accident conditions, fuel cladding tubes made of zirconium alloys can be exposed to steam at high temperature (HT, up 1200 C) before being cooled and then quenched in water. In some conditions, after burst occurrence the cladding can rapidly absorb a significant amount of hydrogen (secondary hydriding), up to 3000 wt.ppm locally, during steam exposition at HT. The study deals with the effect, poorly studied up to date, of high contents of hydrogen on the metallurgical and mechanical properties of two zirconium alloys, Zircaloy-4 and M5, during and after cooling from high temperatures, at which zirconium is in its β phase. A specific facility was developed to homogeneously charge in hydrogen up to ∼ 3000 wt.ppm cladding tube samples of several centimeters in length. Phase transformations, chemical element partitioning and hydrogen precipitation during cooling from the β temperature domain of zirconium were studied by using several techniques, for the materials containing up to ∼ 3000 wt.ppm of hydrogen in average: in-situ neutron diffraction upon cooling from 700 C, X-ray diffraction, μ-ERDA, EPMA and electron microscopy in particular. The results were compared to thermodynamic predictions. In order to study the effect of high hydrogen contents on the mechanical behavior of the (prior-)μ phase of zirconium, axial tensile tests were performed at various temperatures between 20 and 700 C upon cooling from the β temperature domain, on samples with mean hydrogen contents up to ∼ 3000 wt.ppm. The results show that metallurgical and mechanical properties of the (prior-)β phase of zirconium alloys strongly depend on temperature and hydrogen content. (author) [fr

On aging of iron-nickel-titanium alloys

International Nuclear Information System (INIS)

Vintajkin, E.Z.; Dmitriev, V.B.; Udovenko, V.A.

1978-01-01

The mechanism of structural transformations on the initial stages of aging of Fe-(26-29) at. % Ni-(2.5-5.75) at. % Ti alloys was studied by neutron radiography. It was shown that at the earliest aging stages at 550 deg C there appear ordered areas which are FCC nuclei of the Ni 3 Ti phase. The rate of nucleation depends on the content of titanium in the all. In alloys with more than 3% Ti, nuclei appear even at the hardening stage. During the subsequent aging, the nuclei are enriched with nickel and titanium

Nuclear fuel element containing strips of an alloyed Zr, Ti, and Ni getter material

International Nuclear Information System (INIS)

Grossman, L.N.; Packard, D.R.

1975-01-01

A nuclear fuel element for use in the core of a nuclear reactor is disclosed. The nuclear fuel element has disposed therein an alloy having the essential components of nickel, titanium and zirconium, and the alloy reacts with water, water vapor and reactive gases at reactor ambient temperatures. The alloy is disposed in the plenum of the fuel element in the form of strips and preferably the strips are positioned inside a helical member in the plenum. The position of the alloy strips permits gases and liquids entering the plenum to contact and react with the alloy strips. (U.S.)

Segregation in welded nickel-base alloys

International Nuclear Information System (INIS)

Akhtar, J.I.; Shoaib, K.A.; Ahmad, M.; Shaikh, M.A.

1990-05-01

Segregation effects have been investigated in nickel-base alloys monel 400, inconel 625, hastelloy C-276 and incoloy 825, test welded under controlled conditions. Deviations from the normal composition have been observed to varying extents in the welded zone of these alloys. Least effect of this type occurred in Monel 400 where the content of Cu increased in some of the areas. Enhancement of Al and Ti has been found over large areas in the other alloys which has been attributed to the formation of low melting slag. Another common feature is the segregation of Cr, Fe or Ti, most likely in the form of carbides. Enrichment of Al, Ti, Nb, Mb, Mo, etc., to different amounts in some of the areas of these materials is in- terpretted in terms of the formation of gamma prime precipitates or of Laves phases. (author)

In-reactor creep of zirconium alloys by thermal spikes

International Nuclear Information System (INIS)

Ibrahim, E.F.

1975-01-01

The size and duration of thermal spikes from fast neutrons have been calculated for zirconium alloys, showing that spikes up to 1.8 nm radius may exist for 2 x 10 -11 s at greater than melting point, at 570K ambient temperature. Creep rates have been calculated assuming that the elastic strain from the applied stress relaxes in the volume of the spikes (by preferential loop alignment or modification of an existing dislocation network). The calculated rates are consistent with strain rates observed in long term tests-in-reactor, if spike lifetimes are 2 to 2.5 x 10 -11 s. (Auth.)

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

Energy Technology Data Exchange (ETDEWEB)

Alexander, D.J.; Zinkle, S.J.; Rowcliffe, A.F. [Oak Ridge National Lab., TN (United States)

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.

The development of additive manufacturing technique for nickel-base alloys: A review

Science.gov (United States)

Zadi-Maad, Ahmad; Basuki, Arif

2018-04-01

Nickel-base alloys are an attractive alloy due to its excellent mechanical properties, a high resistance to creep deformation, corrosion, and oxidation. However, it is a hard task to control performance when casting or forging for this material. In recent years, additive manufacturing (AM) process has been implemented to replace the conventional directional solidification process for the production of nickel-base alloys. Due to its potentially lower cost and flexibility manufacturing process, AM is considered as a substitute technique for the existing. This paper provides a comprehensive review of the previous work related to the AM techniques for Ni-base alloys while highlighting current challenges and methods to solving them. The properties of conventionally manufactured Ni-base alloys are also compared with the AM fabricated alloys. The mechanical properties obtained from tension, hardness and fatigue test are included, along with discussions of the effect of post-treatment process. Recommendations for further work are also provided.

DC Electric Arc Furnace Application for Production of Nickel-Boron Master Alloys

Science.gov (United States)

Alkan, Murat; Tasyürek, Kerem Can; Bugdayci, Mehmet; Turan, Ahmet; Yücel, Onuralp

2017-09-01

In this study, nickel-boron (Ni-B) alloys were produced via a carbothermic reduction starting from boric acid (H3BO3) with high-purity nickel oxide (NiO), charcoal, and wood chips in a direct current arc furnace. In electric arc furnace experiments, different starting mixtures were used, and their effects on the chemical compositions of the final Ni-B alloys were investigated. After the reduction and melting stages, Ni-B alloys were obtained by tapping from the bottom of the furnace. The samples from the designated areas were also taken and analyzed. The chemical composition of the final alloys and selected samples were measured with wet chemical analysis. The Ni-B alloys had a composition of up to 14.82 mass% B. The phase contents of the final alloys and selected samples were measured using x-ray diffraction (XRD). The XRD data helped predict possible reactions and reaction mechanisms. The material and energy balance calculations were made via the XRD Rietveld and chemical compositions. Nickel boride phases started to form 600 mm below the surface. The targeted NiB phase was detected at the tapping zone of the crucible (850-900 mm depth). The energy consumption was 1.84-4.29 kWh/kg, and the electrode consumption was 10-12 g/kg of raw material charged.

On the rational alloying of structural chromium-nickel steels

International Nuclear Information System (INIS)

Astaf'ev, A.A.

1982-01-01

A study was made on the influence of chromium nickel, phosphorus on the critical brittleness temperature of Cr-Ni-Mo-V structural steels. It is shown that the critical brittleness temperature of these steels increases at chromium content more over than 2% and nickel content more than 2% in the result of carbide transformations during tempering. Increase of nickel content in Cr-Ni-Mo-V-steels strengthens the tendency to embrittlement during slow cooling, from tempering temperature owing to development of process of phosphorus grain-boundary segregation. Two mentioned mechanisms of embrittlement determine principles of rational steel alloying. The extreme dependence of the critical brittleness temperature on chromium and nickel content, which enables to choose the optimum composition of Cr-Ni-Mo-V-steels, was established

Corrosion behaviour of E110- and E635- type zirconium alloys under PWR irradiation simulating conditions

International Nuclear Information System (INIS)

Markelov, V.A.; Novikov, V.V.; Kon'kov, V.F.; Tselishchev, A.V.; Dologov, A.B.; Zmitko, M.; Maserik, V.; Kocik, J.

2008-01-01

As structural materials for VVER 1000 fuel rod claddings and FA components use is made of zirconium alloys E110 (Zr 1Nb) and E635 (Zr 1.2Sn 1Nb 0.35Fe) that meet the design parameters of operation. Nonetheless, the work is in progress to perfect those alloys to reach higher corrosion and shape change resistance. At VNIINM updated E110M and E635M alloys have been developed on E110 and E635 bases. To assess the corrosion behaviour of the updated alloys in comparison to the base alloys their cladding samples were tested in RVS 3 loop of LWR 15 reactor (NRI, Rez) in PWR water chemistry with coolant surface and volume boiling. The data are presented on the influence effected by in pile irradiation for up to 324 days on oxide coat thickness and microstructure of fuel claddings produced from the four tested alloys. It has been revealed that E110 alloy its updated version E110M and E635M alloy compared to E635 have higher corrosion resistances. The paper discusses th+e results on the in pile corrosion of cladding samples from the alloys under study in comparison to the results acquired for similar samples tested in LWR 15 inactive channel and under autoclave conditions. Using methods of TEM, EDX analyses of extraction replicas dislocation structure and phase composition changes were studied in samples of all four alloy claddings LWR 15 reactor irradiated to the material damage dose of 1.5 dpa. The interrelation is discussed between irradiation effected strengthening and corrosion of fuel claddings made of E110 and E635 type zirconium alloys and the evolution of their structure and phase states

Contribution to the study of the electrodeposition of iron-nickel alloys

International Nuclear Information System (INIS)

Valignat, J.

1968-01-01

Using a coulometric technique based upon the anodic intentiostatic dissolution, we studied the potentiostatic, deposition of nickel, iron and nickel iron alloys. We have shown that the minimum of the curve I = f (t) (deposition current versus time) is probably due to the transitory blocking of the surface by hydrogen and that the syn-crystallisation of nickel and iron is responsible for the anomalous co-deposition of these two elements. (author) [fr

Study of the processes for of remelting zirconium alloys in an electric arc furnace

Energy Technology Data Exchange (ETDEWEB)

Pereira, Luiz A.T.; Rossi, Jesualdo L.; Costa, Guilherme R.; Martinez, Luis G.; Sato, Ivone M., E-mail: luiz.atp@uol.com.br, E-mail: jelrossi@ipen.br, E-mail: guilhermeramoscosta@gmail.com, E-mail: lgallego@ipen.br, E-mail: imsato@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

2015-07-01

Zirconium alloy tubes are used as cladding for fuel elements of PWR nuclear reactors, which contains the UO{sub 2} pellets. In the manufacture of these fuel element parts, machining chips from the nuclear grade zirconium alloys are generated. Hence, these machining chips cannot be discarded, as ordinary metallic waste. Thus, the recycling of this material is a strategic aspect for the nuclear technology, both for economic and environmental issues. The main reason is that nuclear grade alloys have very high cost, are not commercially produced in Brazil and has to be imported for the manufacture of the nuclear fuels. This work discusses a method to melt and recycle Zircaloy chips, using an electric-arc furnace to obtain small laboratory ingots. The chemical composition of the ingots was determined using X-ray fluorescence spectroscopy and was compared to the specifications of nuclear grade Zircaloy and to the chemical composition of the received machining chips. The ingots were annealed in high vacuum, as well as were hot rolled in a mill. The microstructures were characterized by optical microscopy. The hardness was evaluated using the Rockwell B scale hardness. The results showed that the compositions of the recycled Zircaloy comply with the chemical specifications and a suitable microstructure has been obtained for nuclear use. (author)

Autoclave Testing on Zirconium Alloy Materials

International Nuclear Information System (INIS)

Hoffmann, Petra-Britt; Sell, Hans-Juergen; Garzarolli, Friedrich

2012-09-01

The corrosion of Zirconium components like fuel rod claddings and spacer grids is limiting lifetime and duty of these components. In Pressurized and Boiling Water Reactors (PWR and BWR), different corrosion phenomena are of interest. Although in-pile experience is the final proof for a material development, significant experience was gained by autoclave tests, trying to simulate in-pile conditions but reducing time for return of experience by increased temperatures. For PWR application, the uniform corrosion is studied in water at up to 370 deg. C and in high pressure steam at 400 deg. C, and for BWR, the nodular corrosion is studied in high pressure steam at 500-520 deg. C. Particular attention has to be given to the corrosion media, because oxidative traces in the water can significantly affect the corrosion response. An extensive air removal is thus important for all corrosion tests. This links to the different water chemistry conditions that have been investigated as separate effects otherwise difficult to separate under in-pile conditions. Uniform corrosion in 350 deg. C water is usually a cyclic process with repeated rate transitions. In addition, at high exposure times an acceleration of corrosion can occur, e.g. for Zr-Sn alloys with a high Sn content. In 400 deg. C steam, corrosion rate decreases somewhat with increasing time. Uniform corrosion rate of Zr alloys depends on their Sn- and Fe+Cr contents as well as on their annealing parameters with a similar trend as in PWR and on their yield strength, however with an opposite trend compared to BWR conditions. Nodular corrosion of BWR alloys depends on the annealing parameter with a similar trend as in PWR and out-of-reactor also significantly on the Fe+Cr content. The hydrogen pickup fraction (HPUF) depends largely on details of the water chemistry and can particularly depend on autoclave degassing and probably also on autoclave contaminations. Thus any HPUF value from out-of- pile corrosion tests is only

Oxidation of zirconium alloys in steam: influence of tetragonal zirconia on oxide growth mechanism

International Nuclear Information System (INIS)

Godlewski, J.

1990-07-01

The oxidation of zirconium alloys in presence of steam, presents after a 'parabolic' growth law, an acceleration of the oxidation velocity. This phenomenon limits the use of zirconium alloys as nuclear fuel cladding element. In order to determine the physico-chemical process leading to this kinetic transition, two approaches have been carried out: the first one has consisted to determine the composition of the oxide layer and its evolution with the oxidation time; and the second one to determine the oxygen diffusion coefficients in the oxide layers of pre- and post-transition as well as their evolution with the oxidation time. The composition of the oxide layers has been determined by two analyses techniques: the X-ray diffraction and the laser Raman spectroscopy. This last method has allowed to confirm the presence of tetragonal zirconium oxide in the oxide layers. Analyses carried out by laser Raman spectroscopy on oxides oblique cuttings have revealed that the tetragonal zirconium oxide is transformed in monoclinic phase during the kinetic transition. A quantitative approach has allowed to corroborate the results obtained by these two techniques. In order to determine the oxygen diffusion coefficients in the oxides layers, two diffusion treatments have been carried out: 1)under low pressure with D 2 18 O 2 ) under high pressure in an autoclave with H 2 18 O. The oxygen 18 concentration profiles have been obtained by two analyses techniques: the nuclear microprobe and the secondary ions emission spectroscopy. The obtained profiles show that the mass transport is made by the volume and particularly by the grain boundaries. The corresponding diffusion coefficients have been calculated with the WHIPPLE and LE CLAIRE solution. The presence of tetragonal zirconium oxide, its relation with the kinetic transition, and the evolution of the diffusion coefficients with the oxidation time, are discussed in terms of internal stresses in the oxide layer and of the oxide layer

Hydrogen-absorbing alloys for the nickel-metal hydride battery

Energy Technology Data Exchange (ETDEWEB)

Mingming Geng; Jianwen Han; Feng Feng [University of Windsor, Ontario (Canada). Mechanical and Materials Engineering; Northwood, D.O. [University of Windsor, Ontario (Canada). Mechanical and Materials Engineering]|[Ryerson Polytechnic University, Toronto (Canada)

1998-12-31

In recent years, owing to the rapid development of portable electronic and electrical appliances, the market for rechargeable batteries has increased at a high rate. The nickel-metal hydride battery (Ni/MH) is one of the more promising types, because of its high capacity, high-rate charge/discharge capability and non-polluting nature. This type of battery uses a hydrogen storage alloy as its negative electrode. The characteristics of the Ni/MH battery, including discharge voltage, high-rate discharge capability and charge/discharge cycle lifetime are mainly determined by the construction of the negative electrode and the composition of the hydrogen-absorbing alloy. The negative electrode of the Ni/MH battery described in this paper was made from a mixture of hydrogen-absorbing alloy, nickel powder and polytetrafluoroethylene (PTFE). A multicomponent MmNi{sub 5}-based alloy (Mm{sub 0.95}Ti{sub 0.05}Ni{sub 3.85} Co{sub 0.45}Mn{sub 0.35}Al{sub 0.35}) was used as the hydrogen-absorbing alloy. The discharge characteristics of the negative electrode, including discharge capacity, cycle lifetime, and polarization overpotential, were studied by means of electrochemical experiments and analysis. The decay of the discharge capacity for the Ni/MH battery (AA size, 1 Ah) was about 1% after 100 charge/discharge cycles and 10% after 500 charge/discharge cycles. (author)

Effect of nickel plating upon tensile tests of uranium--0.75 titanium alloy

International Nuclear Information System (INIS)

Hemperly, V.C.

1975-01-01

Electrolytic-nickel-plated specimens of uranium-0.75 wt percent titanium alloy were tested in air at 20 and 100 percent relative humidities. Tensile-test ductility values were lowered by a high humidity and also by nickel plating alone. Baking the nickel-plated specimens did not eliminate the ductility degradation. Embrittlement because of nickel plating was also evident in tensile tests at -34 0 C. (U.S.)

Anodic characteristics and stress corrosion cracking behavior of nickel rich alloys in bicarbonate and buffer solutions

International Nuclear Information System (INIS)

Zadorozne, Natalia S.; Giordano, Mabel C.; Ares, Alicia E.; Carranza, Ricardo M.; Rebak, Raul B.

2016-01-01

Highlights: • We investigate which element in alloy C-22 may be responsible for the cracking susceptibility of the high nickel alloy. • Six nickel based alloys with different amount of Cr and Mo were selected for the electrochemical tests and response to SSRT. • Polarization tests showed that an anodic peak appear in the passive region in Cr containing alloys. • Cracking of Ni alloys in carbonate solutions seem to be a consequence of the instability of the passivating chromium oxide. • Alloys containing both Cr and Mo have the highest susceptibility. - Abstract: The aim of this work is to investigate which alloying element in C-22 is responsible for the cracking susceptibility of the alloy in bicarbonate and two buffer solutions (tungstate and borate). Six nickel based alloys, with different amount of chromium (Cr) and molybdenum (Mo) were tested using electrochemical methods and slow strain rate tests (SSRT) at 90 °C. All Cr containing alloys had transgranular cracking at high anodic potential; however, C-22 containing high Cr and high Mo was the most susceptible alloy to cracking. Bicarbonate was the most aggressive of three tested environments of similar pH.

Nickel-based gadolinium alloy for neutron adsorption application in ram packages

International Nuclear Information System (INIS)

Robino, C.; McConnell, P.; Mizia, R.

2004-01-01

This paper will outline the results of a metallurgical development program that is investigating the alloying of gadolinium into a nickel-chromium-molybdenum alloy matrix. Gadolinium has been chosen as the neutron absorption alloying element due to its high thermal neutron absorption cross section and low solubility in the expected U.S. repository environment. The nickel-chromium-molybdenum alloy family was chosen for its known corrosion performance, mechanical properties, and weldability. The workflow of this program includes chemical composition definition, primary and secondary melting studies, ingot conversion processes, properties testing, and national consensus codes and standards work. The microstructural investigation of these alloys shows that the gadolinium addition is not soluble in the primary austenite metallurgical phase and is present in the alloy as gadolinium-rich second phase. This is similar to what is observed in a stainless steel alloyed with boron. The mechanical strength values are similar to those expected for commercial Ni-Cr-Mo alloys. The alloys have been corrosion tested in simulated Yucca Mountain aqueous chemistries with acceptable results. The initial results of weldability tests have also been acceptable. Neutronic testing in a moderated critical array has generated favorable results. An American Society for Testing and Materials material specification has been issued for the alloy and a Code Case has been submitted to the American Society of Mechanical Engineers for code qualification. The ultimate goal is acceptance of the alloy for use at the Yucca Mountain repository

Reaction kinetics of oxygen on single-phase alloys, oxidation of nickel and niobium alloys

International Nuclear Information System (INIS)

Lalauze, Rene

1973-01-01

This research thesis first addresses the reaction kinetics of oxygen on alloys. It presents some generalities on heterogeneous reactions (conventional theory, theory of jumps), discusses the core reaction (with the influence of pressure), discusses the influence of metal self-diffusion on metal oxidation kinetics (equilibrium conditions at the interface, hybrid diffusion regime), reports the application of the hybrid diffusion model to the study of selective oxidation of alloys (Wagner model, hybrid diffusion model) and the study of the oxidation kinetics of an alloy forming a solid solution of two oxides. The second part reports the investigation of the oxidation of single phase nickel and niobium alloys (phase α, β and γ)

Aluminium alloys containing iron and nickel; Alliages d'aluminium contenant du fer et du nickel

Energy Technology Data Exchange (ETDEWEB)

Coriou, H.; Fournier, R.; Grall, L.; Hure, J. [Commissariat a l' Energie atomique, Centre d' Etudes Nucleaires de Saclay, Departement de Metallurgie et de Chimie Appliquee (France); Herenguel, J.; Lelong, P. [Centre de Recherches d' Antony, des Trefileries et Laminoirs du Havre (France)

1958-07-01

The first part of this report addresses mechanism, kinetics and structure factors of aluminium alloys containing iron and nickel in water and high temperature steam. The studied alloys contain from 0.3 to 0.7 per cent of iron, and 0.2 to 1.0 per cent of nickel. Corrosion resistance and corrosion structure have been studied. The experimental installation, process and samples are presented. Corrosion structures in water at 350 C are identified and discussed (structure of corrosion products, structure of metal-oxide interface), and then in steam at different temperatures (350-395 C). Corrosion kinetics is experimentally studied (weight variation in time) in water at 350 C and in steam at different temperatures. Reactions occurring at over-heated steam (more than 400 C) are studied, and the case of welded alloys is also addressed. The second part addresses the metallurgical mechanism and processes influencing aluminium alloy resistance to corrosion by high temperature water as it appeared that separated phases protect the solid solution through a neighbourhood action. In order to avoid deep local corrosions, it seems necessary to multiply protective phases in an as uniform as possible way. Some processes enabling this result are described. They belong to conventional metallurgy or to powder metallurgy (with sintering and extrusion)

Assessment of corrosion resistance of cast cobalt- and nickel-chromium dental alloys in acidic environments.

Science.gov (United States)

Mercieca, Sven; Caligari Conti, Malcolm; Buhagiar, Joseph; Camilleri, Josette

2018-01-01

The aim of this study was to compare the degradation resistance of nickel-chromium (Ni-Cr) and cobalt-chromium (Co-Cr) alloys used as a base material for partial dentures in contact with saliva. Wiron® 99 and Wironit Extra-Hard® were selected as representative casting alloys for Ni-Cr and Co-Cr alloys, respectively. The alloys were tested in contact with deionized water, artificial saliva and acidified artificial saliva. Material characterization was performed by X-ray diffractometry (XRD) and microhardness and nanohardness testing. The corrosion properties of the materials were then analyzed using open circuit potential analysis and potentiodynamic analysis. Alloy leaching in solution was assessed by inductively coupled plasma mass spectrometry techniques. Co-Cr alloy was more stable than the Ni-Cr alloy in all solutions tested. Leaching of nickel and corrosion attack was higher in Ni-Cr alloy in artificial saliva compared with the acidified saliva. The corrosion resistance of the Co-Cr alloy was seen to be superior to that of the Ni-Cr alloy, with the former exhibiting a lower corrosion current in all test solutions. Microstructural topographical changes were observed for Ni-Cr alloy in contact with artificial saliva. The Ni-Cr alloy exhibited microstructural changes and lower corrosion resistance in artificial saliva. The acidic changes did not enhance the alloy degradation. Ni-Cr alloys are unstable in solution and leach nickel. Co-Cr alloys should be preferred for clinical use.

An Investigation of the Mechanical Properties of a Weldment of 7% Nickel Alloy Steels

Directory of Open Access Journals (Sweden)

Jeong Yeol Park

2016-11-01

Full Text Available During the last decade, the demand for natural gas has steadily increased for the prevention of environmental pollution. For this reason, many liquefied natural gas (LNG carriers have been manufactured. Since one of the most important issues in the design of LNG carriers is to guarantee structural safety, the use of low-temperature materials is increasing. Among commonly employed low-temperature materials, nickel steel has many benefits such as good strength and outstanding corrosion resistance. Accordingly, nickel steels are one of the most commonly used low-temperature steels for LNG storage tanks. However, the study of fracture toughness with various welding consumables of 7% nickel alloy steel is insufficient for ensuring the structural safety of LNG storage tanks. Therefore, the aim of this study was to evaluate fracture toughness of several different weldments for 7% nickel alloy steels. The weldment of 7% nickel alloy steel was fabricated by tungsten inert gas (TIG, flux cored arc welding (FCAW, and gas metal arc welding (GMAW. In order to assess the material performance of the weldments at low temperature, fracture toughness such as crack tip opening displacement (CTOD and the absorbed impact energy of weldments were compared with those of 9% nickel steel weldments.

Near net shape processing of zirconium or hafnium metals and alloys

International Nuclear Information System (INIS)

Evans, S.C.

1992-01-01

This patent describes a process for producing a metal shape. It comprises: plasma arc melting a metal selected from zirconium, hafnium and alloys thereof comprising at least about 90 w/o of these metals to form a liquid pool; pouring the metal form the pool into a mold to form a near net shape; and reducing the metal from its near net shape to a final size while maintaining the metal temperature below the alpha-beta transition temperature throughout the size reducing step

[The effect of hydrogen peroxide on the electrochemical corrosion properties and metal ions release of nickel-chromium dental alloys].

Science.gov (United States)

Wang, Jue; Qiao, Guang-yan

2013-04-01

To investigate the effect of hydrogen peroxide on the electrochemical corrosion and metal ions release of nickel-chromium dental alloys. The corrosion resistance of nickel-chromium dental alloys was compared by electrochemical impedance spectroscopy (EIS) and potentiodynamic polarization curve (PD) methods in artificial saliva after immersed in different concentrations of hydrogen peroxide for 112 h. The metal ions released from nickel-chromium dental alloys to the artificial saliva were detected after electrochemical measurements using inductively coupled plasma mass spectrometry (ICP-MS). The data was statistically analyzed by analysis of variance (ANOVA) using SPSS 13.0 software package. The electrochemical experiment showed that the sequence of polarization resistance in equivalent circuit (Rct), corrosion potential (Ecorr), pitting breakdown potential (Eb), and the difference between Ecorr and Eb representing the "pseudo-passivation" (δE) of nickel-chromium alloys in artificial saliva was 30% alloys to the artificial saliva, and the order of the concentrations of metal ions was 0% corrosion resistance of nickel-chromium dental alloys decrease after immersed in different concentrations of hydrogen peroxide for 112 h. Nickel-chromium dental alloys are more prone to corrosion in the artificial saliva with the concentration of hydrogen peroxide increased, and more metal ions are released in the artificial saliva.

Effect of deformation and annealing on mechanical properties of nickel-rhenium alloys

International Nuclear Information System (INIS)

Mashkova, V.M.

1978-01-01

Studied have been the mechanical properties of nickel-rhenium alloys, depending on the extent of deformation and heat treatment leading to softening. The mechanical properties of the alloys have been estimated by the results of the tensile tests of wire samples. The softening of the alloy at different temperatures is judged about by the variation in hardness. The results of the study indicate that the most abrupt reduction in the hardness of the cold-hardened metal occurs at 900-1,000 deg C and the hold-time of 1 min. Increase in the hold-time at such temperature almost does not reduce the hardness. It is established that in order to soften nickel-rhenium alloys in the process of the cold-deformation at brief annealings in the air the hold-time should not exceed 5 min at 800-900 deg C

The determination of sulphur in copper, nickel and aluminium alloys by proton activation analysis

International Nuclear Information System (INIS)

Vandecasteele, C.; Dewaele, J.; Esprit, M.; Goethals, P.

1981-01-01

The 34 S(p,n) 34 sup(m)Cl reaction, induced by 13 MeV protons is used for the determination of sulphur in copper, nickel and aluminium alloys. The 34 sup(m)Cl is separated by repeated precipitation as silver chloride. The results obtained were resp. 3.08 +- 0.47, 1.47 +- 0.17 and -1 for copper, nickel and aluminium alloys. (orig.)

Influence of alloying elements on the irradiation hardening and environmental sensitivity of zirconium alloys

International Nuclear Information System (INIS)

Pettersson, K.; Hallstadius, L.; Bergqvist, H.; Nylund, A.; Wikstroem, C.

1992-01-01

Ten different alloys of zirconium have been tested with regard to the effect of irradiation on their mechanical properties and their sensitivity to environmentally induced failure. Two different environments were used: iodine vapour and liquid cesium with an addition of 2% cadmium. The neutron dose was 10 21 n/cm 2 (E>1MeV) and the irradiation temperature was about 300 degrees C. All alloy additions increased the irradiation hardening. Especially notable was the large effect of titanium and tin on irradiation hardening. A limited amount of transmission electron microscopy was carried out in order to find an explanation to the effects. The testing in different environments showed that there is no clear correlation between environmental sensitivity and yield stress. For materials of similar yield stress an alloyed material tends to be more sensitive to environmental cracking than a material which only contains oxygen as an impurity. There also seems to be an effect of oxygen on the environmental cracking sensitivity. A material with 910 ppm oxygen was considerably more sensitive to cracking than a material with 470 ppm oxygen despite the fact that the yield stress values differed by only 90 MPa

Fractography of hydrogen-embrittled iron-chromium-nickel alloys

International Nuclear Information System (INIS)

Caskey, G.R. Jr.

1980-01-01

Tensile specimens of iron-chromium-nickel base alloys were broken in either a hydrogen environment or in air following thermal charging with hydrogen. Fracture surfaces were examined by scanning electron microscopy. Fracture morphology of hydrogen-embrittled specimens was characterized by: changed dimple size, twin-boundary parting, transgranular cleavage, and intergranular separation. The nature and extent of the fracture mode changes induced by hydrogen varied systematically with alloy composition and test temperature. Initial microstructure developed during deformation processing and heat treating had a secondary influence on fracture mode

Phase transformations in intermetallic phases in zirconium alloys

Energy Technology Data Exchange (ETDEWEB)

Filippov, V. P., E-mail: vpfilippov@mephi.ru [National Research Nuclear University MEPhI (Moscow Engineering Physics Institute) (Russian Federation); Kirichenko, V. G. [Kharkiv National Karazin University (Ukraine); Salomasov, V. A. [National Research Nuclear University MEPhI (Moscow Engineering Physics Institute) (Russian Federation); Khasanov, A. M. [University of North Carolina – Asheville, Chemistry Department (United States)

2017-11-15

Phase change was analyzed in intermetallic compounds of zirconium alloys (Zr – 1.03 at.% Fe; Zr – 0.51 at.% Fe; Zr – 0.51 at.% Fe – M(M = Nb, Sn). Mössbauer spectroscopy on {sup 57}Fe nuclei in backscattering geometry with the registration of the internal conversion electrons and XRD were used. Four types of iron bearing intermetallic compounds with Nb were detected. A relationship was found between the growth process of intermetallic inclusions and segregation of these phases. The growth kinetics of inclusions possibly is not controlled by bulk diffusion, and a lower value of the iron atom’s activation energy of migration can be attributed to the existence of enhanced diffusion paths and interface boundaries.

Some observations on the physical metallurgy of nickel alloy weld metals

International Nuclear Information System (INIS)

Skillern, C.G.; Lingenfelter, A.C.

1982-01-01

Numerous nickel alloys play critical roles in various energy-related applications. Successful use of these alloys is almost always dependent on the availability of acceptable welding methods and welding products. An understanding of the physical metallurgy of these alloys and their weld metals and the interaction of weld metal and base metal is essential to take full advantage of the useful properties of the alloys. To illustrate this point, this paper presents data for two materials: INCONEL alloy 718 and INCONEL Welding Electrode 132. 8 figures, 9 tables

Microstructure and age-hardening effects of aluminium alloys with additions of scandium and zirconium

Energy Technology Data Exchange (ETDEWEB)

Galun, R.; Mordike, B.L. [Inst. fuer Werkstoffkunde und Werkstofftechnik, Technische Univ. Clausthal, Clausthal-Zellerfeld (Germany); Maiwald, T.; Smola, B. [Zentrum fuer Funktionswerkstoffe GmbH, Clausthal-Zellerfeld (Germany); Mergen, R.; Manner, M.; Uitz, W. [Miba Gleitlager GmbH, Laakirchen (Australia)

2004-12-01

The aim of the work presented in this report was to produce age-hardenable aluminium alloys containing scandium and zirconium by a casting process with similar cooling conditions like an industrial casting process. Microstructure, precipitation structure and age-hardening response of different alloys with up to 0.4 wt.% Sc and Zr were investigated. Age-hardening experiments from the as-cast condition without solution annealing showed a significant increase of hardness of about 100% for Sc-rich alloys and of 50% for Zr-rich alloys compared to the as-cast condition. TEM investigations revealed the formation of precipitates of ternary Al{sub 3}(Sc{sub x}Zr{sub 1-x}) phases with a cubic cP4 crystal structure. In addition to the strengthening effect, a high thermal stability especially of the precipitates in Zr-rich alloys up to 400 C let these alloys look very promising for high-temperature applications. (orig.)

Dependence of secondary ion emission current on the composition of beryllium-nickel alloys

International Nuclear Information System (INIS)

Pistryak, V.M.; Kozlov, V.F.; Tikhinskij, G.F.; Fogel', Ya.M.

1976-01-01

The dependence is studied of the secondary ions emission current on the composition of beryllium-nickel alloys. It is established that appearance of intermetallide phases in the Be-Ni alloys has no effect on the linear character of the secondary ions Ni + and Be + of emission current. The phase transformation from the solid solution to the compound Ni 5 Be 21 with a change in the alloys concentration is fixed by appearance of the secondary ion NiBe + emission. The limited solubility of nickel in solid beryllium at a temperature close to room temperature is determined to be equal to 1.3+-0.27 at%

Tungsten wire-nickel base alloy composite development

Science.gov (United States)

Brentnall, W. D.; Moracz, D. J.

1976-01-01

Further development and evaluation of refractory wire reinforced nickel-base alloy composites is described. Emphasis was placed on evaluating thermal fatigue resistance as a function of matrix alloy composition, fabrication variables and reinforcement level and distribution. Tests for up to 1,000 cycles were performed and the best system identified in this current work was 50v/o W/NiCrAlY. Improved resistance to thermal fatigue damage would be anticipated for specimens fabricated via optimized processing schedules. Other properties investigated included 1,093 C (2,000 F) stress rupture strength, impact resistance and static air oxidation. A composite consisting of 30v/o W-Hf-C alloy fibers in a NiCrAlY alloy matrix was shown to have a 100-hour stress rupture strength at 1,093 C (2,000 F) of 365 MN/square meters (53 ksi) or a specific strength advantage of about 3:1 over typical D.S. eutectics.

Investigation of Zirconium Oxide Films in Different Dissolved Hydrogen Concentration

International Nuclear Information System (INIS)

Kim, Taeho; Choi, Kyoung Joon; Yoo, Seung Chang; Kim, Ji Hyun

2016-01-01

It has been reported that in pre-transition zirconium oxide, the volume fraction of tetragonal zirconium oxide increased near the oxide/metal (O/M) interface, and the sub-stoichiometric zirconium oxide layer was observed. The diffusion of oxygen ion through the oxide layer is the rate-limiting process during the pre-transition oxidation process, and this diffusion mainly occurs in the grain boundaries. The two layered oxide structure is formed in pre-transition oxide for the zirconium alloy in high-temperature water environment. It is known that the corrosion rate is related to the volume fraction of zirconium oxide and the pores in the oxides; therefore, the aim of this paper is to investigate the oxidation behavior in the pre-transition zirconium oxide in high-temperature water chemistry. In this study, in situ Raman and TEM analysis were conducted for investigating the phase transformation of zirconium alloy in primary water. From this study, the following conclusions are drawn: 1. The zirconium alloy was oxidized in primary water chemistry for 100 d, and Raman and TEM were measured after 30, 50, 80, and 100 d from start-up. 2. TEM and FFT analysis showed that the zirconium oxide mostly consisted of the monoclinic phase. The tetragonal zirconium oxide was just found near the O/M interface

Low cost AB{sub 5}-type hydrogen storage alloys for a nickel-metal hydride battery

Energy Technology Data Exchange (ETDEWEB)

Jiang Lijun [General Res. Inst. for Non-Ferrous Metals, Beijing (China); Zhan Feng [General Res. Inst. for Non-Ferrous Metals, Beijing (China); Bao Deyou [General Res. Inst. for Non-Ferrous Metals, Beijing (China); Qing Guangrong [General Res. Inst. for Non-Ferrous Metals, Beijing (China); Li Yaoquan [General Res. Inst. for Non-Ferrous Metals, Beijing (China); Wei Xiuying [General Res. Inst. for Non-Ferrous Metals, Beijing (China)

1995-12-15

The studies have been carried out on utilizing Ml(NiAl){sub 5}-based alloys as a low cost negative battery electrode. The replacement of nickel by copper improved the cycle lifetime to some extent without a decrease in capacity. Using Ml(NiAlCu){sub 5} alloys, hydrogen storage alloys with good overall characteristics and low cost were obtained through substituting cobalt or silicon for nickel. The discharge capacity was further increased by increasing the lanthanum content in lanthanum-rich mischmetal. (orig.)

The Development of an In-Situ TEM Technique for Studying Corrosion Behavior as Applied to Zirconium-Based Alloys

Science.gov (United States)

Harlow, Wayne

Zirconium-based alloys are a commonly used material for nuclear fuel rod cladding, due to its low neutron cross section and good corrosion properties. However, corrosion is still a limiting factor in fuel rod lifespan, which restricts burn up levels, and thus efficiency, that can be achieved. While long-term corrosion behavior has been studied through both reactor and autoclave samples, the oxide nucleation and growth behavior has not been extensively studied. This work develops a new technique to study the initial stages of corrosion in zirconium-based alloys and the microstructural effects on this process by developing an environmental cell system for the TEM. Nanoscale oxidation parameters are developed, as is a new FIB technique to support this method. Precession diffraction is used in conjunction with in-situ TEM to observe the initial stages of corrosion in these alloys, and oxide thickness is estimated using low-loss EELS. In addition, the stress stabilization of tetragonal ZrO 2 is explored in the context of sample preparation for TEM. It was found that in-situ environmental TEM using an environmental cell replicates the oxidation behavior observed in autoclaved samples in both oxide structure and phases. Utilizing this technique, it was shown that cracking of the oxide layer in zirconium-based alloys is related to oxide relaxation, and not thermal changes. The effect of secondary phase particles on oxidation behavior did not present significant results, however a new method for studying initial oxidation rates using low-loss EELS was developed.

Intermetallic Nickel-Titanium Alloys for Oil-Lubricated Bearing Applications

Science.gov (United States)

DellaCorte, C.; Pepper, S. V.; Noebe, R.; Hull, D. R.; Glennon, G.

2009-01-01

An intermetallic nickel-titanium alloy, NITINOL 60 (60NiTi), containing 60 wt% nickel and 40 wt% titanium, is shown to be a promising candidate material for oil-lubricated rolling and sliding contact applications such as bearings and gears. NiTi alloys are well known and normally exploited for their shape memory behavior. When properly processed, however, NITINOL 60 exhibits excellent dimensional stability and useful structural properties. Processed via high temperature, high-pressure powder metallurgy techniques or other means, NITINOL 60 offers a broad combination of physical properties that make it unique among bearing materials. NITINOL 60 is hard, electrically conductive, highly corrosion resistant, less dense than steel, readily machined prior to final heat treatment, nongalling and nonmagnetic. No other bearing alloy, metallic or ceramic encompasses all of these attributes. Further, NITINOL 60 has shown remarkable tribological performance when compared to other aerospace bearing alloys under oil-lubricated conditions. Spiral orbit tribometer (SOT) tests were conducted in vacuum using NITINOL 60 balls loaded between rotating 440C stainless steel disks, lubricated with synthetic hydrocarbon oil. Under conditions considered representative of precision bearings, the performance (life and friction) equaled or exceeded that observed with silicon nitride or titanium carbide coated 440C bearing balls. Based upon this preliminary data, it appears that NITINOL 60, despite its high titanium content, is a promising candidate alloy for advanced mechanical systems requiring superior and intrinsic corrosion resistance, electrical conductivity and nonmagnetic behavior under lubricated contacting conditions.

Oxide characterization and hydrogen behaviors of Zr-based alloys

International Nuclear Information System (INIS)

Kim, Y. S.; Kim, D. J.; Kwon, S. H.; Lee, H. S.; Oh, S. J.; Yim, B. J.; Son, S. B.; Yun, S. P.

2006-03-01

The work scope and contents of the research are as follows : basic properties of zirconium alloys, hydrogen pick-up mechanism of zirconium alloy, effects of hydride on the corrosion behaviors of zirconium alloys, estimation on stress of oxide layer in the zirconium alloy, microstructure and characteristic of oxide in pre-hydrided zirconium alloys

Aluminium-nickel-iron alloys resistant to corrosion by water at high temperature. Their basic properties - their improvement

International Nuclear Information System (INIS)

Coriou, H.; Fournier, R.; Grall, L.; Hure, J.

1959-01-01

The development of the investigations carried out on these alloys is reviewed, showing the establishment of their fundamental, particularly structural, properties. This is followed by studies on: 1 - The penetration process in corrosion. The results of micrographic studies of the metal oxide interface are given for a series of alloys treated in water and steam between 350 and 395 deg. C. The hypothesis of attack by pockets of gas pressure is corroborated, and a second process of deep penetration by islands of intergranular-type corrosion is shown to take place. These patches, distinct from the surface corrosion layer and sometimes forming at a considerable depth inside the metal, would be due to heterogeneities in composition of the solid solution making up the matrix of these alloys. 2 - The role of titanium and zirconium additions on rolled metal. Systematic studies are carried out on a series of alloys with titanium and zirconium contents between 0.05 and 0.15 per cent. The favourable effect of titanium in particular has been demonstrated. Zirconium acts in the same way, but less efficiently. The improvement due to these additions can be compared to their action on the distribution of the second phases, which tend to become more pronounced and more homogeneously distributed. The influence of solder on these alloys has been studied, showing up the part played by the structure gradients introduced by fission. (author) [fr

Anisotropy of mechanical properties of zirconium and zirconium alloys

International Nuclear Information System (INIS)

Medrano, R.E.

1975-01-01

In studies of technological applications of zirconium to fuel elements of nuclear reactor, it was found that the use of plasticity equations for isotropic materials is not in agreement with experimental results, because of the strong anisotropy of zirconium. The present review describes recent progress on the knowledge of the influence of anisotropy on mechanical properties, after Douglass' review in 1971. The review was written to be selfconsistent, changing drastically the presentation of some of the referenced papers. It is also suggested some particular experiments to improve developments in this area

Effect of Electric Voltage and Current of X-ray Chamber on the Element inthe Zirconium Alloy Analysis X-ray by X-ray Fluorescence

International Nuclear Information System (INIS)

Yusuf-Nampira; Narko-Wibowo, L; Rosika-Krisnawati; Nudia-Barenzani

2000-01-01

The using of x-ray fluorescence in the chemical analysis depend heavilyon the parameters of x-ray chamber, for examples : electric voltage andelectric current. That parameter give effect in the result of determine ofSn, Cr, Fe and Ni in the zirconium alloy. 20 kV electric voltages are used onthe Mo x-ray chamber shall product x-ray of zirconium in the sample materialcan give effect in the stability of the analysis result (deviation more than5%). The result of analysis of elements in the zirconium alloy shall givedeviation less than 5% when using of electric voltage of the x-ray chamberless than 19 kV. The sensitivity of analysis can be reached by step upelectric current of x-ray chamber. (author)

Pitting morphologies of zirconium base alloys in aqueous and non aqueous chloride media

International Nuclear Information System (INIS)

Palit, G.C.; Gadiyar, H.S.

1988-01-01

Pitting morphology of zirconium and Zr-Cr alloys in aqueous chloride and nonaqueous methanol + 0.4 per cent HCl solution was investigated and observed to follow different modes in these two environments. While in aqueous chloride solution pitting was transgranular and randomly oriented, in methanol-chloride solution pits were observed to initiate and propagate along the grain boundaries. In aqueous chloride solution very irregular and sponge like zirconium metal was formed inside the pit while in methanol-chloride solution the pits were crystallographic in nature. Optical microscopy has revealed that pits preferentially initiate and propagate along scratch line in aqueous chloride solution, but such was not the case in nonaqueous methanol-chloride solution. The nature and the mechanism operating in the catastropic failure of these materials are investigated. (author). 10 refs., 11 figs

Influence of the alloying effect on nickel K-shell fluorescence yield in Ni Si alloys

Science.gov (United States)

Kalayci, Y.; Agus, Y.; Ozgur, S.; Efe, N.; Zararsiz, A.; Arikan, P.; Mutlu, R. H.

2005-02-01

Alloying effects on the K-shell fluorescence yield ωK of nickel in Ni-Si binary alloy system have been studied by energy dispersive X-ray fluorescence. It is found that ωK increases from pure Ni to Ni 2Si and then decreases from Ni 2Si to NiSi. These results are discussed in terms of d-occupation number on the Ni site and it is concluded that electronic configuration as a result of p-d hybridization explain qualitatively the observed variation of ωK in Ni-Si alloys.

Collaborative analysis for certification of zirconium and zirconium base alloy reference materials JAERI-Z11 to Z16

International Nuclear Information System (INIS)

1985-03-01

The second Sub-Committee on Zircaloy Analysis was organized in April 1978, under the Committee on Analytical Chemistry on Nuclear Fuels and Reactor Materials, JAERI, for the renewal of zirconium and zirconium base alloy certified reference materials (CRMs). The Sub-Committee carried out collaborative analysis among 13 participating laboratories for the certification of the CRMs, JAERI-Z11 to Z18, after development, improvement and evaluation of analytical methods during the period of May 1978 to June 1982. As the result of the collaborative analysis, the certified value was given for 18 elements (Sn, Fe, Ni, Cr, B, Cd, U, Cu, Co, Mn, Pb, Al, Ti, Si, Mo, W, Hf, C) in the CRMs. The first part of this report includes general discussion, the second part principles of certification, the third part development and verification of analytical methods, and the fourth part evaluation of analytical results on 17 elements. Preparation of Z11 to Z18, and certification for carbon in JAERI-Z17 and Z18 were reported separately in JAERI-M 83-241 and M 83-035, respectively. (author)

Synthesis of Complex-Alloyed Nickel Aluminides from Oxide Compounds by Aluminothermic Method

Directory of Open Access Journals (Sweden)

Victor Gostishchev

2018-06-01

Full Text Available This paper deals with the investigation of complex-alloyed nickel aluminides obtained from oxide compounds by aluminothermic reduction. The aim of the work was to study and develop the physicochemical basis for obtaining complex-alloyed nickel aluminides and their application for enhancing the properties of coatings made by electrospark deposition (ESD on steel castings, as well as their use as grain refiners for tin bronze. The peculiarities of microstructure formation of master alloys based on the Al–TM (transition metal system were studied using optical, electronic scanning microscopy and X-ray spectral microanalysis. There were regularities found in the formation of structural components of aluminum alloys (Ni–Al, Ni-Al-Cr, Ni-Al-Mo, Ni-Al-W, Ni-Al-Ti, Ni-Cr-Mo-W, Ni-Al-Cr-Mo-W-Ti, Ni-Al-Cr-V, Ni-Al-Cr-V-Mo and changes in their microhardness, depending on the composition of the charge, which consisted of oxide compounds, and on the amount of reducing agent (aluminum powder. It is shown that all the alloys obtained are formed on the basis of the β phase (solid solution of alloying elements in nickel aluminide and quasi-eutectic, consisting of the β′ phase and intermetallics of the alloying elements. The most effective alloys, in terms of increasing microhardness, were Al-Ni-Cr-Mo-W (7007 MPa and Al-Ni-Cr-V-Mo (7914 MPa. The perspective is shown for applying the synthesized intermetallic master alloys as anode materials for producing coatings by electrospark deposition on steel of C1030 grade. The obtained coatings increase the heat resistance of steel samples by 7.5 times, while the coating from NiAl-Cr-Mo-W alloy remains practically nonoxidized under the selected test conditions. The use of NiAl intermetallics as a modifying additive (0.15 wt. % in tin bronze allows increasing the microhardness of the α-solid solution by 1.9 times and the microhardness of the eutectic (α + β phase by 2.7 times.

Properties of experimental copper-aluminium-nickel alloys for dental post-and-core applications.

Science.gov (United States)

Rittapai, Apiwat; Urapepon, Somchai; Kajornchaiyakul, Julathep; Harniratisai, Choltacha

2014-06-01

This study aimed to develop a copper-aluminium-nickel alloy which has properties comparable to that of dental alloys used for dental post and core applications with the reasonable cost. Sixteen groups of experimental copper alloys with variants of 3, 6, 9, 12 wt% Al and 0, 2, 4, 6 wt% Ni were prepared and casted. Their properties were tested and evaluated. The data of thermal, physical, and mechanical properties were analyzed using the two-way ANOVA and Tukey's test (α=0.05). The alloy toxicity was evaluated according to the ISO standard. The solidus and liquidus points of experimental alloys ranged from 1023℃ to 1113℃ and increased as the nickel content increased. The highest ultimate tensile strength (595.9 ± 14.2 MPa) was shown in the Cu-12Al-4Ni alloy. The tensile strength was increased as the both elements increased. Alloys with 3-6 wt% Al exhibited a small amount of 0.2% proof strength. Accordingly, the Cu-9Al-2Ni and Cu-9Al-4Ni alloys not only demonstrated an appropriate modulus of elasticity (113.9 ± 8.0 and 122.8 ± 11.3 GPa, respectively), but also had a value of 0.2% proof strength (190.8 ± 4.8 and 198.2 ± 3.4 MPa, respectively), which complied with the ISO standard requirement (>180 MPa). Alloys with the highest contents of nickel (6 wt% Ni) revealed a widespread decolourisation zone (5.0-5.9 mm), which correspondingly produced the largest cell response, equating positive control. The copper alloys fused with 9 wt% Al and 2-4 wt% Ni can be considered for a potential use as dental post and core applications.

The solidification velocity of nickel and titanium alloys

Science.gov (United States)

Altgilbers, Alex Sho

2002-09-01

The solidification velocity of several Ni-Ti, Ni-Sn, Ni-Si, Ti-Al and Ti-Ni alloys were measured as a function of undercooling. From these results, a model for alloy solidification was developed that can be used to predict the solidification velocity as a function of undercooling more accurately. During this investigation a phenomenon was observed in the solidification velocity that is a direct result of the addition of the various alloying elements to nickel and titanium. The additions of the alloying elements resulted in an additional solidification velocity plateau at intermediate undercoolings. Past work has shown a solidification velocity plateau at high undercoolings can be attributed to residual oxygen. It is shown that a logistic growth model is a more accurate model for predicting the solidification of alloys. Additionally, a numerical model is developed from simple description of the effect of solute on the solidification velocity, which utilizes a Boltzmann logistic function to predict the plateaus that occur at intermediate undercoolings.

Hydride formation on deformation twin in zirconium alloy

Energy Technology Data Exchange (ETDEWEB)

Kim, Ju-Seong [Korea Atomic Energy Research Institute, 989-111 Daedeokdaero, Yuseong-gu, Daejeon, 305-353 (Korea, Republic of); Kim, Sung-Dae [Korea Institute of Material Science (KIMS), 797 Changwondaero, Changwon, Gyeongnam, 642-831 (Korea, Republic of); Yoon, Jonghun, E-mail: yooncsmd@gmail.com [Department of Mechanical Engineering, Hanyang University, 1271 Sa3-dong, Sangrok-gu, Ansan-si, Gyeonggi-do, 426-791 (Korea, Republic of)

2016-12-15

Hydrides deteriorate the mechanical properties of zirconium (Zr) alloys used in nuclear reactors. Intergranular hydrides that form along grain boundaries have been extensively studied due to their detrimental effects on cracking. However, it has been little concerns on formation of Zr hydrides correlated with deformation twins which is distinctive heterogeneous nucleation site in hexagonal close-packed metals. In this paper, the heterogeneous precipitation of Zr hydrides at the twin boundaries was visualized using transmission electron microscopy. It demonstrates that intragranular hydrides in the twinned region precipitates on the rotated habit plane by the twinning and intergranular hydrides precipitate along the coherent low energy twin boundaries independent of the conventional habit planes. Interestingly, dislocations around the twin boundaries play a substantial role in the nucleation of Zr hydrides by reducing the misfit strain energy.

The effect of substrate texture and oxidation temperature on oxide texture development in zirconium alloys

Energy Technology Data Exchange (ETDEWEB)

Garner, A., E-mail: alistair.garner@manchester.ac.uk [Materials Performance Centre, University of Manchester, Grosvenor Street, Manchester, M17HS (United Kingdom); Frankel, P. [Materials Performance Centre, University of Manchester, Grosvenor Street, Manchester, M17HS (United Kingdom); Partezana, J. [Westinghouse Electric Company, 1332 Beulah Road, Pittsburgh, PA 15235 (United States); Preuss, M. [Materials Performance Centre, University of Manchester, Grosvenor Street, Manchester, M17HS (United Kingdom)

2017-02-15

During corrosion of zirconium alloys a highly textured oxide is formed, the degree of this preferred orientation has previously been shown to be an important factor in determining the corrosion behaviour of these alloys. Two distinct experiments were designed in order to investigate the origin of this oxide texture development on two commercial alloys. Firstly, sheet samples of Zircaloy-4 were oxidised between 500 and 800 °C in air. The resulting monoclinic oxide texture strength was observed to decrease with increasing oxidation temperature. In a second experiment, orthogonal faces of Low Tin ZIRLO{sub ™} were oxidised in 360 °C water, providing different substrate textures but identical microstructures. The substrate texture was observed to have a negligible effect on the corrosion performance whilst the major orientation of both oxide phases was found to be independent of substrate orientation. It is concluded that the main driving force for oxide texture development in single-phase zirconium alloys is the compressive stress caused by the Zr−ZrO{sub 2} transformation. - Highlights: • Substrate orientation does not significantly affect oxide texture development. • Corrosion performance is independent of substrate texture. • Monoclinic oxide texture strength decreases with increasing oxidation temperature. • The main driving force for texture development is the oxidation-induced stress.

Corrosion behaviour of cladded nickel base alloys

International Nuclear Information System (INIS)

Brandl, W.; Ruczinski, D.; Nolde, M.; Blum, J.

1995-01-01

As a consequence of the high cost of nickel base alloys their use as surface layers is convenient. In this paper the properties of SA-as well as RES-cladded NiMo 16Cr16Ti and NiCr21Mo14W being produced in single and multi-layer technique are compared and discussed with respect to their corrosion behaviour. Decisive criteria describing the qualities of the claddings are the mass loss, the susceptibility against intergranular corrosion and the pitting corrosion resistance. The results prove that RES cladding is the most suitable technique to produce corrosion resistant nickel base coatings. The corrosion behaviour of a two-layer RES deposition shows a better resistance against pitting than a three layer SAW cladding. 7 refs

Process and equipement for zone heat treatment of zirconium alloys tubes

International Nuclear Information System (INIS)

Kiesler, A.J.; Frischmann, P.G.; Rockwood, A.C.

1977-01-01

Process for the thermal treatment of an area of a long zirconium alloy part in order to increase its corrosion resistance in the cooling conditions of boiling water reactor, in which the part is moved lengthwise through a succession of critical maximum temperature areas, during a critical time and is subjected to a temperature reduction at critical rate, so that each successive portion reaches a maximum temperature between 825 0 C, and directing water at a temperature around 60 to 80 0 C as jets in the cooling area [fr

Inconel type resistive alloys based on ultrahigh purity nickel

International Nuclear Information System (INIS)

Matsarin, K.A.; Matsarin, S.K.

2000-01-01

The new nickel high-ohm alloys (ρ = 1.2-1.4 μOhm · m), containing the W, Al, Mo alloying elements in the quantity, not exceeding their solubility in a solid solution, are developed on the basis of the Inconel-type standard alloy. The optical composition of the alloy was determined by the results of the alloy was determined by the results of the electric resistance measurement and technological effectiveness indices (relative to the pressure and workable metal yield). The following optimal component concentrations were established: 14-17 %Cr; 10-12 %Fe; 0.5-1.0 %Cu; 1.0-1.5 %Mn; 0.1-0.2 %C; 0.4-0.6 %Si; 0.5-3.0 %W; 5-16 %Mo; 0.5-2.0 %Al; the remainder - Ni. The new alloys are recommended as materials for resistive elements of direct-glow cathode nodes of low capacity electron tubes [ru

Void swelling and segregation in dilute nickel alloys

International Nuclear Information System (INIS)

Potter, D.I.; Rehn, L.E.; Okamoto, P.R.; Wiedersich, H.

1977-01-01

Five binary alloys containing 1 at.% of Al, Ti, Mo, Si and Be in nickel were irradiated at temperatures from 525 to 675 0 C with 3.5-MeV 58 Ni + ions. The resultant microstructures were examined by TEM, and void diameters, number densities and swelling are presented for each alloy over the temperature interval investigated. A systematic relation between solute misfit (size factor) and void swelling is established for these alloys. Solute concentration profiles near the irradiated surface were determined and these also exhibited a systematic behavior--undersize solutes segregated to the surface, whereas oversize solutes were depleted. The results are consistent with calculations based on strong interstitial-solute trapping by undersize solutes and vacancy-solute trapping by oversize solutes that are weak interstitial traps

Fabrication of tungsten wire reinforced nickel-base alloy composites

Science.gov (United States)

Brentnall, W. D.; Toth, I. J.

1974-01-01

Fabrication methods for tungsten fiber reinforced nickel-base superalloy composites were investigated. Three matrix alloys in pre-alloyed powder or rolled sheet form were evaluated in terms of fabricability into composite monotape and multi-ply forms. The utility of monotapes for fabricating more complex shapes was demonstrated. Preliminary 1093C (2000F) stress rupture tests indicated that efficient utilization of fiber strength was achieved in composites fabricated by diffusion bonding processes. The fabrication of thermal fatigue specimens is also described.

Galvanic corrosion resistance of welded dissimilar nickel-base alloys

International Nuclear Information System (INIS)

Corbett, R.A.; Morrison, W.S.; Snyder, R.J.

1986-01-01

A program for evaluating the corrosion resistance of various dissimilar welded nickel-base alloy combinations is outlined. Alloy combinations included ALLCORR, Hastelloy C-276, Inconel 72 and Inconel 690. The GTAW welding process involved both high and minimum heat in-put conditions. Samples were evaluated in the as-welded condition, as well as after having been aged at various condtions of time and temperature. These were judged to be most representative of process upset conditions which might be expected. Corrosion testing evaluated resistance to an oxidizing acid and a severe service environment in which the alloy combinations might be used. Mechanical properties are also discussed

Evolution of Nickel-titanium Alloys in Endodontics.

Science.gov (United States)

Ounsi, Hani F; Nassif, Wadih; Grandini, Simone; Salameh, Ziad; Neelakantan, Prasanna; Anil, Sukumaran

2017-11-01

To improve clinical use of nickel-titanium (NiTi) endodontic rotary instruments by better understanding the alloys that compose them. A large number of engine-driven NiTi shaping instruments already exists on the market and newer generations are being introduced regularly. While emphasis is being put on design and technique, manufacturers are more discreet about alloy characteristics that dictate instrument behavior. Along with design and technique, alloy characteristics of endodontic instruments is one of the main variables affecting clinical performance. Modification in NiTi alloys is numerous and may yield improvements, but also drawbacks. Martensitic instruments seem to display better cyclic fatigue properties at the expense of surface hardness, prompting the need for surface treatments. On the contrary, such surface treatments may improve cutting efficiency but are detrimental to the gain in cyclic fatigue resistance. Although the design of the instrument is vital, it should in no way cloud the importance of the properties of the alloy and how they influence the clinical behavior of NiTi instruments. Dentists are mostly clinicians rather than engineers. With the advances in instrumentation design and alloys, they have an obligation to deal more intimately with engineering consideration to not only take advantage of their possibilities but also acknowledge their limitations.

The development of hydrogen storage electrode alloys for nickel hydride batteries

Science.gov (United States)

Hong, Kuochih

The development of hydrogen storage electrode alloys in the 1980s resulted in the birth and growth of the rechargeable nickel hydride (Ni/MH) battery. In this paper we describe briefly a semi-empirical electrochemical/thermodynamic approach to develop/screen a hydrogen storage alloy for electrochemical application. More specifically we will discuss the AB x Ti/Zr-based alloys. Finally, the current state of the Ni/MH batteries including commercial manufacture processes, cell performance and applications is given.

Formation of chemical compounds under vacuum plasma-arc deposition of nickel and its alloy onto piezoceramics

International Nuclear Information System (INIS)

Grinchenko, V.T.; Lyakhovich, T.K.; Prosina, N.I.; Khromov, S.M.

1988-01-01

The phase composition of the transition layer appearing during vacuum-arc coating of nickel and nickel alloy with copper on barium titanate and lead zirconate-titanate is identified. During vacuum plasma-arc coating of nickel and its alloy at the boundary with barium titanate and lead zirconate-titanate the Ni 2 Ti 4 O compound appears which has the crystal lattice type identical with substrate with the parity of lattice parameters. The transition layer contains nickel oxides and NiTiO 3 in the case of barium titanate. When titanate content in substrate increases the zone of reaction diffusion increases in value and becomes more complicate in composition

Microstructural modelling and lubrication study during zirconium alloy hot extrusion

International Nuclear Information System (INIS)

Gaudout, B.

2009-01-01

Using torsion tests (with strain rate jumps) and an experimental hot mini-extrusion apparatus, several samples zirconium alloy have been deformed: Zircaloy-4 (high α range) and Zr-1Nb (α + β domain). The fragmentation of the microstructure and post-dynamic grain growth have been examined. The main difference between these two alloys is that Zr-1Nb does not show grain growth during a heat treatment within the α + β domain after hot deformation. The recrystallization volume fraction has been measured on extruded samples with or without heat treatment. These rheological and microstructural data have been used to determine the parameters of a microstructural model including: a work-hardening model (Laaasraoui/Jonas), a continuous dynamic recrystallization model (Gourdet/Montheillet) and a grain growth model. This model leads to a good prediction of recrystallization volume fraction for Zircaloy-4 extrusion. However, the Zr-1Nb model cannot be validated because of the difficulty to observe deformed microstructures. Extrusion process is lubricated with a solid film. Trapping tests show that this lubricant is thermoviscoplastic. Friction along the container and several observations show the lubrication is not realized by a continuous film. Indeed, the heterogeneousness of deformation of these alloys causes a rupture of the lubricant film. Experiments and numerical simulations show that the radial gradient of axial displacement is affected by friction but also by stress softening of the alloys. (author)

Evolution of zirconium-based precipitates during oxidation and irradiation of Zr alloys (impact on the oxidation kinetics of Zr alloys)

International Nuclear Information System (INIS)

Pecheur, Dominique

1993-01-01

As the oxidation of the zircaloy sheath is one of the factors which limit the lifetime of nuclear fuel rods, this research thesis aims at a better knowledge of the involved oxidation mechanisms and to improve the oxidation resistance in order to increase rod lifetime. Oxidation test performed in autoclave to study zirconium alloy oxidation without irradiation showed that oxidation kinetics is significantly higher under irradiation. This difference is attributed to a different evolution of the sheath material under irradiation. Thus, this research focused on the role of precipitates in the oxidation process of zirconium alloys, and on the impact of their amorphization on this oxidation. After a detailed description of the context and of the various implemented experimental means, the author presents the results obtained on a reference material on the one hand, and on a material irradiated by ions or neutrons on the other hand. More particularly, the author studied in these both cases the introduction of precipitates in the oxide layer by transmission electronic microscopy, and oxidation kinetics obtained in autoclave on these two types of material. He reports the analysis of the introduction of precipitates in the oxide layer formed on the reference material. He proposes interpretations for the evolutions of structure and of chemical compositions of precipitates in the oxide layer. These observations are then correlated with oxidation kinetics in these alloys. Finally, the author discusses results of oxidation tests obtained on materials irradiated by ions and by neutrons [fr

In vitro and in vivo corrosion evaluation of nickel-chromium- and copper-aluminum-based alloys.

Science.gov (United States)

Benatti, O F; Miranda, W G; Muench, A

2000-09-01

The low resistance to corrosion is the major problem related to the use of copper-aluminum alloys. This in vitro and in vivo study evaluated the corrosion of 2 copper-aluminum alloys (Cu-Al and Cu-Al-Zn) compared with a nickel-chromium alloy. For the in vitro test, specimens were immersed in the following 3 corrosion solutions: artificial saliva, 0.9% sodium chloride, and 1.0% sodium sulfide. For the in vivo test, specimens were embedded in complete dentures, so that one surface was left exposed. The 3 testing sites were (1) close to the oral mucosa (partial self-cleaning site), (2) surface exposed to the oral cavity (self-cleaning site), and (3) specimen bottom surface exposed to the saliva by means of a tunnel-shaped perforation (non-self-cleaning site). Almost no corrosion occurred with the nickel-chromium alloy, for either the in vitro or in vivo test. On the other hand, the 2 copper-aluminum-based alloys exhibited high corrosion in the sulfide solution. These same alloys also underwent high corrosion in non-self-cleaning sites for the in vivo test, although minimal attack was observed in self-cleaning sites. The nickel-chromium alloy presented high resistance to corrosion. Both copper-aluminum alloys showed considerable corrosion in the sulfide solution and clinically in the non-self-cleaning site. However, in self-cleaning sites these 2 alloys did not show substantial corrosion.

Alkaline stress corrosion of iron-nickel-chromium austenitic alloys

International Nuclear Information System (INIS)

Hocquellet, Dominique

1984-01-01

This research thesis reports the study of the behaviour in stress corrosion of austenitic iron-nickel-chromium alloys by means of tensile tests at imposed strain rate, in a soda solution at 50 pc in water and 350 degrees C. The author shows that the mechanical-chemical model allows the experimental curves to be found again, provided the adjustment of characteristic parameters, on the one hand, of corrosion kinetics, and on the other hand, of deformation kinetics. A classification of the studied alloys is proposed [fr

Advanced nickel base alloys for high strength, corrosion applications

Science.gov (United States)

Flinn, J.E.

1998-11-03

Improved nickel-base alloys of enhanced strength and corrosion resistance, produced by atomization of an alloy melt under an inert gas atmosphere and of composition 0--20Fe, 10--30Cr, 2--12Mo, 6 max. Nb, 0.05--3 V, 0.08 max. Mn, 0.5 max. Si, less than 0.01 each of Al and Ti, less than 0.05 each of P and S, 0.01--0.08C, less than 0.2N, 0.1 max. 0, bal. Ni. 3 figs.

Hydrogen permeation inhibition by zinc-nickel alloy plating on steel XC68

International Nuclear Information System (INIS)

El Hajjami, A.; Gigandet, M.P.; De Petris-Wery, M.; Catonne, J.C.; Duprat, J.J.; Thiery, L.; Raulin, F.; Starck, B.; Remy, P.

2008-01-01

The inhibition of hydrogen permeation and barrier effect by zinc-nickel plating was investigated using the Devanathan-Stachurski permeation technique. The hydrogen permeation and hydrogen diffusion for the zinc-nickel (12-15%) plating on steel XC68 is compared with zinc and nickel. Hydrogen permeation and hydrogen diffusion were followed as functions of time at current density applied (cathodic side) and potential permanent (anodic side). The hydrogen permeation inhibition for zinc-nickel is intermediate to that of nickel and zinc. This inhibition was due to nickel-rich layer effects at the Zn-Ni alloy/substrate interface, is shown by GDOES. Zinc-nickel plating inhibited the hydrogen diffusion greater as compared to zinc. This diffusion resistance was due to the barrier effect caused by the nickel which is present at the interface and transformed the hydrogen atomic to Ni 2 H compound, as shown by GIXRD.

Hydrogen permeation inhibition by zinc-nickel alloy plating on steel XC68

Energy Technology Data Exchange (ETDEWEB)

El Hajjami, A. [Institut UTINAM, UMR CNRS 6213, Sonochimie et Reactivite des Surfaces, Universite de Franche-Comte, 16 route de Gray, 25030 Besancon Cedex (France); Coventya S.A.S., 51 rue Pierre, 92588 Clichy Cedex (France); Gigandet, M.P. [Institut UTINAM, UMR CNRS 6213, Sonochimie et Reactivite des Surfaces, Universite de Franche-Comte, 16 route de Gray, 25030 Besancon Cedex (France)], E-mail: marie-pierre.gigandet@univ-fcomte.fr; De Petris-Wery, M. [Institut Universitaire de Technologie d' Orsay, Universite Paris XI, Plateau de Moulon, 91400 Orsay (France); Catonne, J.C. [Professeur Honoraire du Conservatoire national des arts et metiers (CNAM), Paris (France); Duprat, J.J.; Thiery, L.; Raulin, F. [Coventya S.A.S., 51 rue Pierre, 92588 Clichy Cedex (France); Starck, B.; Remy, P. [Lisi Automotive, 28 faubourg de Belfort, BP 19, 90101 Delle Cedex (France)

2008-12-30

The inhibition of hydrogen permeation and barrier effect by zinc-nickel plating was investigated using the Devanathan-Stachurski permeation technique. The hydrogen permeation and hydrogen diffusion for the zinc-nickel (12-15%) plating on steel XC68 is compared with zinc and nickel. Hydrogen permeation and hydrogen diffusion were followed as functions of time at current density applied (cathodic side) and potential permanent (anodic side). The hydrogen permeation inhibition for zinc-nickel is intermediate to that of nickel and zinc. This inhibition was due to nickel-rich layer effects at the Zn-Ni alloy/substrate interface, is shown by GDOES. Zinc-nickel plating inhibited the hydrogen diffusion greater as compared to zinc. This diffusion resistance was due to the barrier effect caused by the nickel which is present at the interface and transformed the hydrogen atomic to Ni{sub 2}H compound, as shown by GIXRD.

Nickel coating on high strength low alloy steel by pulse current deposition

Science.gov (United States)

Nigam, S.; Patel, S. K.; Mahapatra, S. S.; Sharma, N.; Ghosh, K. S.

2015-02-01

Nickel is a silvery-white metal mostly used to enhance the value, utility, and lifespan of industrial equipment and components by protecting them from corrosion. Nickel is commonly used in the chemical and food processing industries to prevent iron from contamination. Since the properties of nickel can be controlled and varied over broad ranges, nickel plating finds numerous applications in industries. In the present investigation, pulse current electro-deposition technique has been used to deposit nickel on a high strength low alloy (HSLA) steel substrate.Coating of nickel is confirmed by X-ray diffraction (XRD) and EDAX analysis. Optical microscopy and SEM is used to assess the coating characteristics. Electrochemical polarization study has been carried out to study the corrosion behaviour of nickel coating and the polarisation curves have revealed that current density used during pulse electro-deposition plays a vital role on characteristics of nickel coating.

High-temperature thermodynamic activities of zirconium in platinum alloys determined by nitrogen-nitride equilibria

International Nuclear Information System (INIS)

Goodman, D.A.

1980-05-01

A high-temperature nitrogen-nitride equilibrium apparatus is constructed for the study of alloy thermodynamics to 2300 0 C. Zirconium-platinum alloys are studied by means of the reaction 9ZrN + 11Pt → Zr 9 Pt 11 + 9/2 N 2 . Carful attention is paid to the problems of diffusion-limited reaction and ternary phase formation. The results of this study are and a/sub Zr//sup 1985 0 C/ = 2.4 x 10 -4 in Zr 9 Pt 11 ΔG/sub f 1985 0 C/ 0 Zr 9 Pt 11 less than or equal to -16.6 kcal/g atom. These results are in full accord with the valence bond theory developed by Engel and Brewer; this confirms their prediction of an unusual interaction of these alloys

Anodic behaviours, dissolution and passivation of iron-nickel alloys in sulphuric environment. Influence of friction

International Nuclear Information System (INIS)

Ponthiaux, Pierre

1990-01-01

This research thesis reports the study of anodic dissolution and passivation of iron-nickel alloys (10, 20 and 31 pc nickel) in a sulphuric environment, with or without friction, by using anodic polarization curves. Without friction, the three alloys have a similar behaviour as pure iron. The analysis reveals different dissolution and passivation mechanisms with pure iron, and highlights the influence of nickel content on corresponding kinetics. The influence of cyclic plane-on-plane friction has been studied for the 31 pc nickel alloy which has an unsteady austenitic structure. Fretting results in some modifications of polarization curves. These modifications are analysed with respect to fretting parameters (relative speed of antagonist surfaces, contact pressure). They reveal the specific influence of the following phenomena: material strain hardening, martensitic transformation induced by strain hardening, partial destruction of adsorbates and/or of the passive film. Modifications of polarization curves give also information on the evolution of friction characteristics with respect to speed (a phenomenon of lubrication by the electrolyte occurs) [fr

Thermal creep behavior of N36 zirconium alloy cladding tube

International Nuclear Information System (INIS)

Wang, P.; Zhao, W.; Dai, X.

2015-01-01

N36 is an alloy containing Zr, Sn, Nb and Fe that is developed by China as a superior cladding material to meet the performance of PWR fuel assembly at the maximum fuel rod burn-up. The creep characteristics of N36 zirconium alloy cladding tube were investigated at temperature from 593 K to 723 K with stress ranging from 20 MPa to 160 MPa. Transitions in creep mechanisms were noted, showing the distinct three rate-controlled creep mechanisms for the alloy at test conditions. In the region of low stresses with stress exponent n ∼ 1 and activation energy Q ∼ (104±4) kJ.mol -1 , Coble creep, based on diffusion of materials through grain boundaries, is the dominant rate-controlling mechanism, which contributes to the creep deformation. The formation of slip bands acts as an accommodation mechanism. In the region of middle stress with stress exponent n ∼ 3 and activation energy Q ∼ (195±7) kJ.mol -1 , micro-creep, caused by viscous gliding of dislocations due to the interaction of O atoms with dislocations, controls the deformation. In the high stress region with stress exponent n ∼ 5-6 and activation energy Q ∼ (210±10) kJ.mol -1 , two mechanisms of the climb of edge dislocations (EDC) and the motion of jogged screw dislocation (MJS) contribute to rate controlling process. In test conditions N36 alloy cladding tube behaves a type of creep similar to that noted in class-I (A) alloys

Heterogeneous coarsening of Pb phase and the effect of Cu addition on it in a nanophase composite of Al-10 wt%Pb alloy prepared by mechanical alloying

International Nuclear Information System (INIS)

Zhu, M.; Liu, X.; Wu, Z.F.; Ouyang, L.Z.; Zeng, M.Q.

2009-01-01

A nanophase composite of Al-10 wt%Pb alloy was prepared by mechanical alloying. The coarsening behavior of Pb phase in the composite during heating process was investigated by X-ray diffraction, scanning electron microscopy, transmission electron microscopy, and nanoindentation test. The present work shows that the Pb phase grew substantially and had two different size distributions when the heating temperature was above 823 K. The different size distributions of Pb phase were owing to different grain size ranges of Al matrix in different regions, which led to the different growth rates of the Pb phase in those regions. It has been proposed that the different size ranges of Al grain appeared upon heating were originated from a statistical size distribution of Al grains in the as-milled powder. With the addition of a small amount of Cu, the heterogeneous growth of Pb phase can be suppressed, and the coarsening of Pb phase shows two distinct rates. This indicates that the coarsening is mainly governed by grain boundary diffusion and lattice diffusion of Al matrix in the initial stage and the later one, respectively

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

International Nuclear Information System (INIS)

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.

Problems of zirconium metal production in Czechoslovakia

International Nuclear Information System (INIS)

Vareka, J.; Vaclavik, E.

1975-01-01

The problems are summed up of the production and quality control of zirconium sponge. A survey is given of industrial applications of zirconium in form of pure metal or alloys in nuclear power production, ferrous and non-ferrous metallurgy, chemical engineering and electrical engineering. A survey is also presented of the manufacture of zirconium metal in advanced capitalist countries. (J.B.)

Corrosion Behavior and Oxide Properties of Zr-Nb-Cu and Zr-Nb-Sn Alloy in High Dissolved Hydrogen Primary Water Chemistry

Energy Technology Data Exchange (ETDEWEB)

Lee, Yun Ju; Kim, Tae Ho; Kim, Ji Hyun [UNIST, Ulsan (Korea, Republic of)

2016-05-15

The water-metal interface is regarded as rate-controlling site governing the rapid oxidation transition in high burn-up fuel. And the zirconium oxide is made in water-metal interface and its structure and phase do an important role in terms of oxide properties. During oxidation process, the protective tetragonal oxide layer develops at the interface due to accumulated high stress during oxide growth, and it turns into non-protective monoclinic oxide with increasing oxide thickness, thus decreasing the stress. It has been reported that Nb addition was proven to be very beneficial for increasing the corrosion resistance of the zirconium alloys. From a more recent study, Cu addition in Nb containing Zirconium alloy was reported to be effective for increasing corrosion resistance in water containing B and Li. According to the previous research conducted, Zr-Nb-Cu shows better corrosion resistance than Zircaloy-4. The dissolved hydrogen (DH) concentration is the key issue of primary water chemistry, and the effect of DH concentration on the corrosion rate of nickel based alloy has been researched. However, the effect of DH on the zirconium alloy corrosion mechanism was not fully investigated. In this study, the weight gain measurement, FIB-SEM analysis, and Raman spectroscopic measurement were conducted to investigate the effects of dissolved hydrogen concentration and the chemical composition on the corrosion resistance and oxide phase of Zr-Nb-Cu alloy and Zr-Nb-Sn alloy after oxidizing in a primary water environment for 20 d. The corrosion rate of Zr-Nb-Cu alloy is slow, when it is compared to Zr-Nb-Sn alloy. In SEM images, the oxide thickness of Zr-Nb-Cu alloy is measured to be around 1.06 μm it of Zr-Nb-Sn alloy is measured to be 1.15 μm. It is because of the Segregation made by Sn solute element when Sn solute element oxidized. And according to ex situ Raman spectra, Zr-Nb-Cu alloy oxide has more tetragonal zirconium oxide fraction than Zr-Nb-Sn alloy oxide.

The degradation of zirconium alloys in nuclear reactors - a review

International Nuclear Information System (INIS)

Lim, D.; Graham, N.A.

1986-01-01

This report presents the findings of a survey of available non-Canadian literature on the oxidation and hydriding of zirconium alloys. Much of the literature was found to address the Zircaloys, particularly when used as fuel cladding subjected to a radioactive and oxidizing environment. Hydriding of Zircaloys is mainly attributed to oxidation. The survey revealed that Zr-Nb alloys have been included in some investigations; however, data on the long-term degradation of Zr-2.5 wt% Nb, in particular, were scarce. The reviewed literature did not lead to conclusions regarding the potential for accelerated hydriding due to corrosion at crevices and/or second-phase particles, nor did it lead to conclusions as to the potential for a 'breakaway' in oxidation and hydrogen acquisition in long service life of Zr-Nb alloys. Specific information on service experience in U.S.S.R. power reactors could not be obtained; however, most of the information surveyed leads to the conclusion that fuel channels having Zr-2.5 wt% Nb pressure tubes should perform satisfactorily with respect to degradation from corrosion and hydriding provided they are installed correctly and are not operated under conditions that are far removed from those anticipated in design. 91 refs

[The effect of epigallocatechin gallate (EGCG) on the surface properties of nickel-chromium dental casting alloys after electrochemical corrosion].

Science.gov (United States)

Qiao, Guang-yan; Zhang, Li-xia; Wang, Jue; Shen, Qing-ping; Su, Jian-sheng

2014-08-01

To investigate the effect of epigallocatechin gallate (EGCG) on the surface properties of nickel-chromium dental alloys after electrochemical corrosion. The surface morphology and surface structure of nickel-chromium dental alloys were examined by stereomicroscope and scanning electron microscopy before and after electrochemical tests in 0 g/L and 1.0 g/L EGCG artificial saliva. The surface element component and chemical states of nickel-chromium dental alloys were analyzed by X-ray photoelectron spectrograph after electrochemical tests in 0 g/L and 1.0 g/L EGCG artificial saliva. More serious corrosion happened on the surface of nickel-chromium alloy in 1.0 g/L EGCG artificial saliva than in 0 g/L EGCG. The diameters of corrosion pits were smaller, and the dendrite structure of the alloy surface was not affected in 0 g/L EGCG. While the diameters of corrosion pits were larger, the dendritic interval of the alloy surface began to merge, and the dendrite structure was fuzzy in 1.0 g/L EGCG. In addition, the O, Ni, Cr, Be, C and Mo elements were detected on the surface of nickel-chromium alloys after sputtered for 120 s in 0 g/L EGCG and 1.0 g/L EGCG artificial saliva after electrochemical corrosion, and the surface oxides were mainly NiO and Cr(2)O(3). Compared with 0 g/L EGCG artificial saliva, the content of O, NiO and Cr(2)O(3) were lower in 1.0 g/L EGCG. The results of surface morphology and the corrosion products both show that the corrosion resistance of nickel-chromium alloys become worse and the oxide content of corrosion products on the surface reduce in 1.0 g/L EGCG artificial saliva.

Solidification Mapping of a Nickel Alloy 718 Laboratory VAR Ingot

Science.gov (United States)

Watt, Trevor J.; Taleff, Eric M.; Lopez, Felipe; Beaman, Joe; Williamson, Rodney

The solidification microstructure of a laboratory-scale Nickel alloy 718 vacuum arc remelted (VAR) ingot was analyzed. The cylindrical, 210-mm-diameter ingot was sectioned along a plane bisecting it length-wise, and this mid-plane surface was ground and etched using Canada's reagent to reveal segregation contrast. Over 350 photographs were taken of the etched mid-plane surface and stitched together to form a single mosaic image. Image data in the resulting mosaic were processed using a variety of algorithms to extract quantities such as primary dendrite orientation, primary dendrite arm spacing (PDAS), and secondary dendrite arm spacing (SDAS) as a function of location. These quantities were used to calculate pool shape and solidification rate during solidification using existing empirical relationships for Nickel Alloy 718. The details and outcomes of this approach, along with the resulting comparison to experimental processing conditions and computational models, are presented.

Nuclear fuel element containing particles of an alloyed Zr, Ti, and Ni getter material

International Nuclear Information System (INIS)

Grossman, L.N.; Levin, H.A.

1975-01-01

A nuclear fuel element for use in the core of a nuclear reactor is disclosed. The nuclear fuel element has disposed therein an alloy having the essential components of nickel, titanium and zirconium, and the alloy reacts with water, water vapor and reactive gases at reactor ambient temperatures. The alloy is disposed in the plenum of the fuel element in the form of particles in a hollow gas permeable container having a multiplicity of openings of size smaller than the size of the particles. The openings permit gases and liquids entering the plenum to contact the particles of alloy. The container is preferably held in the spring in the plenum of the fuel element. (Official Gazette)

Determination of hydrogen in uranium-niobium-zirconium alloy by inert-gas fusion

International Nuclear Information System (INIS)

Carden, W.F.

1979-12-01

An improved method has been developed using inert-gas fusion for determining the hydrogen content in uranium-niobium-zirconium (U-7.5Nb-2.5Zr) alloy. The method is applicable to concentrations of hydrogen ranging from 1 to 250 micrograms per gram and may be adjusted for analysis of greater hydrogen concentrations. Hydrogen is determined using a hydrogen determinator. The limit of error for a single determination at the 95%-confidence level (at the 3.7-μg/g-hydrogen level) is +-1.4 micrograms per gram hydrogen

X-ray study of texture in zirconium alloy tubes and in graphite

International Nuclear Information System (INIS)

Skvortsov, V.V.; Alekseev, S.I.

1987-01-01

X-ray study of texture in zirconium alloy tubes and in graphite has been developed. The method is based on constructing coordinate grid of stereographic projection determining quantity and coordinates of points where measurements should be performed depending on a specimen slope pitch. Complete stereographic projection obtained so is a base both for constructing pole figures showing distribution normales of plane system being studied and for calculating texture coefficients determining property anisotropy in materials under investigation. This method can be applied to study texture in items of any materials independent of the item shape

Effect of zirconium addition on the recrystallization behaviour of a ...

Indian Academy of Sciences (India)

In the present work, zirconium was added to a commercial Al–Cu–Mg alloy and by heat treatment Al3Zr particles were precipitated and after forging, the grain size was an order of magnitude lower than the alloy without zirconium. Transmission electron microscopy was employed to characterize the second phase particles, ...

Method of fabricating thin-walled articles of tungsten-nickel-iron alloy

Science.gov (United States)

Hovis, Jr., Victor M.; Northcutt, Jr., Walter G.

1982-01-01

The present invention relates to a method for fabricating thin-walled high-density structures oftungsten-nickel-iron alloys. A powdered blend of the selected alloy constituents is plasma sprayed onto a mandrel having the desired article configuration. The sprayed deposit is removed from the mandrel and subjected to liquid phase sintering to provide the alloyed structure. The formation of the thin-walled structure by plasma spraying significantly reduces shrinkage, and cracking while increasing physical properties of the structure over that obtainable by employing previously known powder metallurgical procedures.

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

Science.gov (United States)

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.

Corrosion investigation of multilayered ceramics and experimental nickel alloys in SCWO process environments

International Nuclear Information System (INIS)

Garcia, K.M.; Mizia, R.

1995-02-01

A corrosion investigation was done at MODAR, Inc., using a supercritical water oxidation (SCWO) vessel reactor. Several types of multilayered ceramic rings and experimental nickel alloy coupons were exposed to a chlorinated cutting oil TrimSol, in the SCWO process. A corrosion casing was designed and mounted in the vessel reactor with precautions to minimize chances of degrading the integrity of the pressure vessel. Fifteen of the ceramic coated rings were stacked vertically in the casing at one time for each test. There was a total of 36 rings. The rings were in groupings of three rings that formed five sections. Each section saw a different SCWO environment, ranging from 650 to 300 degrees C. The metal coupons were mounted on horizontal threaded holders welded to a vertical rod attached to the casing cover in order to hang down the middle of the casing. The experimental nickel alloys performed better than the baseline nickel alloys. A titania multilayered ceramic system sprayed onto a titanium ring remained intact after 120-180 hours of exposure. This is the longest time any coating system has withstood such an environment without significant loss

Corrosion investigation of multilayered ceramics and experimental nickel alloys in SCWO process environments

Energy Technology Data Exchange (ETDEWEB)

Garcia, K.M.; Mizia, R.

1995-02-01

A corrosion investigation was done at MODAR, Inc., using a supercritical water oxidation (SCWO) vessel reactor. Several types of multilayered ceramic rings and experimental nickel alloy coupons were exposed to a chlorinated cutting oil TrimSol, in the SCWO process. A corrosion casing was designed and mounted in the vessel reactor with precautions to minimize chances of degrading the integrity of the pressure vessel. Fifteen of the ceramic coated rings were stacked vertically in the casing at one time for each test. There was a total of 36 rings. The rings were in groupings of three rings that formed five sections. Each section saw a different SCWO environment, ranging from 650 to 300{degrees}C. The metal coupons were mounted on horizontal threaded holders welded to a vertical rod attached to the casing cover in order to hang down the middle of the casing. The experimental nickel alloys performed better than the baseline nickel alloys. A titania multilayered ceramic system sprayed onto a titanium ring remained intact after 120-180 hours of exposure. This is the longest time any coating system has withstood such an environment without significant loss.

Characterisation of hydrides in a zirconium alloy, by EBSD

International Nuclear Information System (INIS)

Ubhi, H.S.; Larsen, K.

2012-01-01

Zirconium alloys are used in nuclear reactors owing to their low capture cross-section for thermal neutrons and good mechanical and corrosion properties. However, they do suffer from delayed hydrogen cracking (DHC) due to formation of hydride particles. This study shows how the electron back-scatter diffraction (EBSD) technique can be used to characterise hydrides and their orientation relationship with the matrix. Hydrided EB weld specimens were prepared by electro-polishing, characterised using Oxford instruments AZtecHKL EBSD apparatus and software attached to a FEG SEM. Hydrides were found to exist as fine intra granular plates and having the Blackburn orientation relationship, i.e. (0002)Zr//(111)hydride and (1120)Zr//(1-10)hydride. The hydrides were also found to contain sigma 3 boundaries as well as local misorientations. (author)

Preparation, characterization and wear behavior of carbon coated magnesium alloy with electroless plating nickel interlayer

International Nuclear Information System (INIS)

Mao, Yan; Li, Zhuguo; Feng, Kai; Guo, Xingwu; Zhou, Zhifeng; Dong, Jie; Wu, Yixiong

2015-01-01

Highlights: • The carbon film with nickel interlayer (Ni + C coating) is deposited on GW83. • In Ni + C composite coating the carbon coating has good adhesion with the nickel interlayer. • The wear track of Ni + C coating is narrower compared to the bare one. • The wear resistance of GW83 is greatly improved by the Ni + C coating. - Abstract: Poor wear resistance of rare earth magnesium alloys has prevented them from wider application. In this study, composite coating (PVD carbon coating deposited on electroless plating nickel interlayer) is prepared to protect GW83 magnesium alloys against wear. The Ni + C composite coating has a dense microstructure, improved adhesion strength and hardness due to the effective support of Ni interlayer. The wear test result shows that the Ni + C composite coating can greatly prolong the wear life of the magnesium alloy. The wear track of the Ni + C coated magnesium alloy is obviously narrower and shows less abrasive particles as compared with the bare one. Abrasive wear is the wear mechanism of the coatings at the room temperature. In conclusion, the wear resistance of the GW83 magnesium alloy can be greatly improved by the Ni + C composite coating

Thermal Coefficient of Linear Expansion Modified by Dendritic Segregation in Nickel-Iron Alloys

Science.gov (United States)

Ogorodnikova, O. M.; Maksimova, E. V.

2018-05-01

The paper presents investigations of thermal properties of Fe-Ni and Fe-Ni-Co casting alloys affected by the heterogeneous distribution of their chemical elements. It is shown that nickel dendritic segregation has a negative effect on properties of studied invars. A mathematical model is proposed to explore the influence of nickel dendritic segregation on the thermal coefficient of linear expansion (TCLE) of the alloy. A computer simulation of TCLE of Fe-Ni-Co superinvars is performed with regard to a heterogeneous distribution of their chemical elements over the whole volume. The ProLigSol computer software application is developed for processing the data array and results of computer simulation.

Process for etching zirconium metallic objects

International Nuclear Information System (INIS)

Panson, A.J.

1988-01-01

In a process for etching of zirconium metallic articles formed from zirconium or a zirconium alloy, wherein the zirconium metallic article is contacted with an aqueous hydrofluoric acid-nitric acid etching bath having an initial ratio of hydrofluoric acid to nitric acid and an initial concentration of hydrofluoric and nitric acids, the improvement, is described comprising: after etching of zirconium metallic articles in the bath for a period of time such that the etching rate has diminished from an initial rate to a lesser rate, adding hydrofluoric acid and nitric acid to the exhausted bath to adjust the concentration and ratio of hydrofluoric acid to nitric acid therein to a value substantially that of the initial concentration and ratio and thereby regenerate the etching solution without removal of dissolved zirconium therefrom; and etching further zirconium metallic articles in the regenerated etching bath

Iron-nickel alloys as canister material for radioactive waste disposal in underground repositories

International Nuclear Information System (INIS)

Apps, J.A.

1982-01-01

Canisters containing high-level radioactive waste must retain their integrity in an underground waste repository for at least one thousand years after burial (Nuclear Regulatory Commission, 1981). Since no direct means of verifying canister integrity is plausible over such a long period, indirect methods must be chosen. A persuasive approach is to examine the natural environment and find a suitable material which is thermodynamically compatible with the host rock under the environmental conditions with the host rock under the environmental conditions expected in a waste repository. Several candidates have been proposed, among them being iron-nickel alloys that are known to occur naturally in altered ultramafic rocks. The following review of stability relations among iron-nickel alloys below 350 0 C is the initial phase of a more detailed evaluation of these alloys as suitable canister materials

Surface preparation process of a uranium titanium alloy, in particular for chemical nickel plating

International Nuclear Information System (INIS)

Henri, A.; Lefevre, D.; Massicot, P.

1987-01-01

In this process the uranium alloy surface is attacked with a solution of lithium chloride and hydrochloric acid. Dissolved uranium can be recovered from the solution by an ion exchange resin. Treated alloy can be nickel plated by a chemical process [fr

Hydrogen storage in metallic hydrides: the hydrides of magnesium-nickel alloys

International Nuclear Information System (INIS)

Silva, E.P. da.

1981-01-01

The massive and common use of hydrogen as an energy carrier requires an adequate solution to the problem of storing it. High pressure or low temperatures are not entirely satisfactory, having each a limited range of applications. Reversible metal hydrides cover a range of applications intermediate to high pressure gas and low temperature liquid hydrogen, retaining very favorable safety and energy density characteristics, both for mobile and stationary applications. This work demonstrates the technical viability of storing hydrogen in metal hydrides of magnesium-nickel alloys. Also, it shows that technology, a product of science, can be generated within an academic environment, of the goal is clear, the demand outstanding and the means available. We review briefly theoretical models relating to metal hydride properties, specially the thermodynamics properties relevant to this work. We report our experimental results on hydrides of magnesium-nickel alloys of various compositions including data on structure, hydrogen storage capacities, reaction kinetics, pressure-composition isotherms. We selected a promising alloy for mass production, built and tested a modular storage tank based on the hydrides of the alloy, with a capacity for storing 10 Nm sup(3) of hydrogen of 1 atm and 20 sup(0)C. The tank weighs 46,3 Kg and has a volume of 21 l. (author)

Nickel-Titanium Alloys: Corrosion "Proof" Alloys for Space Bearing, Components and Mechanism Applications

Science.gov (United States)

DellaCorte, Christopher

2010-01-01

An intermetallic nickel-titanium alloy, 60NiTi (60 wt% Ni, 40 wt% Ti), is shown to be a promising candidate tribological material for space mechanisms. 60NiTi offers a broad combination of physical properties that make it unique among bearing materials. 60NiTi is hard, electrically conductive, highly corrosion resistant, readily machined prior to final heat treatment, and is non-magnetic. Despite its high Ti content, 60NiTi is non-galling even under dry sliding. No other bearing alloy, metallic or ceramic, encompasses all of these attributes. Since 60NiTi contains such a high proportion of Ti and possesses many metallic properties, it was expected to exhibit poor tribological performance typical of Ti alloys, namely galling type behavior and rapid lubricant degradation. In this poster-paper, the oil-lubricated behavior of 60NiTi is presented.

In situ corrosion testing of various nickel alloys at Måbjerg waste incineration plant

DEFF Research Database (Denmark)

Montgomery, Melanie; Hansson, A. N.; Jensen, S. A.

2013-01-01

overlay material currently being used to give improved corrosion resistance. In order to assess the use of alternative nickel alloys, test panels have been manufactured and inserted into Måbjerg waste incineration plant. Inconel 625 as a 50% weld overlay, two layered weld overlay and as a spiral weld......The majority of waste in Denmark is disposed via waste to energy (WTE) incineration plants which are fabricated from carbon steel. However, due to the increasing corrosiveness of waste over the years, more corrosion resistant alloys are required. In Denmark, Inconel 625 (UNSN06625) is the weld...... overlay was exposed. Other nickel materials exposed were weld overlay Alloy 686, Alloy 50 and Sumitomo Super 625 coextruded tube. Exposure has been undertaken from 2003 to 2009 in the first pass and 2005–2009 in the second pass, and sections have been removed and investigated during this period...

Effective and Environmentally Friendly Nickel Coating on the Magnesium Alloy

Directory of Open Access Journals (Sweden)

Ivana Škugor Rončević

2016-12-01

Full Text Available The low density and good mechanical properties make magnesium and its alloys attractive construction materials in the electronics, automotive, and aerospace industry, together with application in medicine due to their biocompatibility. Magnesium AZ91D alloy is an alloy with a high content of aluminum, whose mechanical properties overshadow the low corrosion resistance caused by the composition of the alloy and the existence of two phases: α magnesium matrix and β magnesium aluminum intermetallic compound. To improve the corrosion resistance, it is necessary to find an effective protection method for the alloy surface. Knowing and predicting electrochemical processes is an essential for the design and optimization of protective coatings on magnesium and its alloys. In this work, the formations of nickel protective coatings on the magnesium AZ91D alloy surface by electrodeposition and chemical deposition, are presented. For this purpose, environmentally friendly electrolytes were used. The corrosion resistance of the protected alloy was determined in chloride medium using appropriate electrochemical techniques. Characterization of the surface was performed with highly sophisticated surface-analytical methods.

Analysis of nickel-base alloys by Grimm-type glow discharge emission and x-ray fluorescence spectrometry

International Nuclear Information System (INIS)

Ferreira, N.P.; Strauss, J.A.; Van Maarseveen, I.; Ivanfy, A.B.

1985-01-01

Nickel-base alloys can be analysed as satisfactorily as steels by XRF as well as by the Grimm-type source, in spite of problems caused by element combinations, spectral line overlap and the influence of the structure and heat conduction properties on sputtering in the glow discharge source. This extended abstract briefly discusses the use of Grimm-type glow discharge emission and XRF as techniques for the analysis of nickel-base alloys

Design of a single variable helium effects experiment for irradiation in FFTF [Fast Flux Test Facility] using alloys enriched in nickel 59

International Nuclear Information System (INIS)

Simons, R.L.; Brager, H.R.; Matsumoto, W.Y.

1986-03-01

Nickel enriched in nickel 59 was extracted from the fragments of a fracture toughness specimen of Inconel 600 irradiated in the Engineering Test Reactor (ETR). The nickel contained 2.0% nickel 59. Three heats of austenitic steel doped with nickel-59 were prepared and inserted in the Materials Open Test Assembly (MOTA) of the Fast Flux Test Facility (FFTF). The experiment was single variable in helium effects because chemically identical alloys without nickel-59 were being irradiated side by side with the doped material. The alloys doped with nickel 59 produced 10 to 100 times more helium than the control alloys. The materials included ternary and quaternary alloys in the form of transmission electron microscope (TEM) discs and miniature tensile specimens. The helium to dpa ratio was in the range 5 to 35 and was nearly constant throughout the irradiation. The exposures ranged from 0.25 to 50 displacements per atom (dpa) over the duration of the experiment. The irradiation temperatures covered the range of 360 to 600 0 C

Structural evolution in films of alloy Zn70Al27Cu3 (ZA27)

International Nuclear Information System (INIS)

Zhu, Y.H.; Lee, W.B.; Mei, Z.; To, S.; Sze, Y.K.

2005-01-01

Films of alloy ZA27 were produced using electron deposition technique. Structural evolution and phase decomposition of the films were studied. It was found that the alloy films were relatively stable because of a strong preferred crystal orientation of the nano-phases. The dependence of nano-phase stability on the Zn content and the preferred crystal orientation is discussed from point of view of Gibbs free energy

Dependency of Delayed Hydride Crack Velocity on the Direction of an Approach to Test Temperatures in Zirconium Alloys

International Nuclear Information System (INIS)

Kim, Young Suk; Kim, Kang Soo; Im, Kyung Soo; Ahn, Sang Bok; Cheong, Yong Moo

2005-01-01

Recently, Kim proposed a new DHC model where a driving force for the DHC is a supersaturated hydrogen concentration as a result of a hysteresis of the terminal solid solubility (TSS) of hydrogen in zirconium alloys upon a heating and a cooling. This model was demonstrated to be valid through a model experiment where the prior plastic deformation facilitated nucleation of the reoriented hydrides, thus reducing the supersaturated hydrogen concentration at the plastic zone ahead of the crack tip and causing hydrogen to move to the crack tip from the bulk region. Thus, an approach to the test temperature by a cooling is required to create a supersaturation of hydrogen, which is a driving force for the DHC of zirconium alloys. However, despite the absence of the supersaturation of hydrogen due to an approach to the test temperature by a heating, DHC is observed to occur in zirconium alloys at the test temperatures below 180 .deg. C. As to this DHC phenomenon, Kim proposed that stress-induced transformation from γ-hydrides to δ-hydrides is likely to be a cause of this, based on Root's observation that the γ-hydride is a stable phase at temperatures lower than 180 .deg. C. In other words, the hydrides formed at the crack tip would be δ-hydrides due to the stressinduced transformation while the bulk region still maintains the initial hydride phase or γ-hydrides. It should be noted that Ambler has also assumed the crack tip hydrides to be δ-hydrides. When the δ-hydrides or ZrH1.66 are precipitated at the crack tip due to the transformation of the γ-hydrides or ZrH, the crack tip will have a decreased concentration of dissolved hydrogen in zirconium, considering the atomic ratio of hydrogen and zirconium in the γ- and δ-hydrides. In contrast, due to no stress-induced transformation of hydrides, the bulk region maintains the initial concentration of dissolved hydrogen. Hence, there develops a difference in the hydrogen concentration or .C between the bulk and the

Dependency of Delayed Hydride Crack Velocity on the Direction of an Approach to Test Temperatures in Zirconium Alloys

Energy Technology Data Exchange (ETDEWEB)

Kim, Young Suk; Kim, Kang Soo; Im, Kyung Soo; Ahn, Sang Bok; Cheong, Yong Moo [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

2005-07-01

Recently, Kim proposed a new DHC model where a driving force for the DHC is a supersaturated hydrogen concentration as a result of a hysteresis of the terminal solid solubility (TSS) of hydrogen in zirconium alloys upon a heating and a cooling. This model was demonstrated to be valid through a model experiment where the prior plastic deformation facilitated nucleation of the reoriented hydrides, thus reducing the supersaturated hydrogen concentration at the plastic zone ahead of the crack tip and causing hydrogen to move to the crack tip from the bulk region. Thus, an approach to the test temperature by a cooling is required to create a supersaturation of hydrogen, which is a driving force for the DHC of zirconium alloys. However, despite the absence of the supersaturation of hydrogen due to an approach to the test temperature by a heating, DHC is observed to occur in zirconium alloys at the test temperatures below 180 .deg. C. As to this DHC phenomenon, Kim proposed that stress-induced transformation from {gamma}-hydrides to {delta}-hydrides is likely to be a cause of this, based on Root's observation that the {gamma}-hydride is a stable phase at temperatures lower than 180 .deg. C. In other words, the hydrides formed at the crack tip would be {delta}-hydrides due to the stressinduced transformation while the bulk region still maintains the initial hydride phase or {gamma}-hydrides. It should be noted that Ambler has also assumed the crack tip hydrides to be {delta}-hydrides. When the {delta}-hydrides or ZrH1.66 are precipitated at the crack tip due to the transformation of the {gamma}-hydrides or ZrH, the crack tip will have a decreased concentration of dissolved hydrogen in zirconium, considering the atomic ratio of hydrogen and zirconium in the {gamma}- and {delta}-hydrides. In contrast, due to no stress-induced transformation of hydrides, the bulk region maintains the initial concentration of dissolved hydrogen. Hence, there develops a difference in the

The behaviour of zirconium alloys in Santowax OM organic coolant at high temperatures

International Nuclear Information System (INIS)

Sawatzky, A.

1964-10-01

Zirconium alloys have been exposed to Santowax OM at temperatures of 320 to 400 o C for times as long as 5000 hours. Short-term experiments (less than 2 weeks) were done in stainless-steel bombs and small out-of-pile loops. The X-7 organic loop in the NRX reactor was used to study long-term oxidation and hydriding both in-flux and out-of-flux. The results obtained lead to several tentative conclusions: Aluminum cladding serves as an effective hydrogen barrier; Considerable protection against hydriding is given by zirconium oxide, provided impurities in the organic are carefully controlled; Hydriding is greatly enhanced by the presence of chlorine in the coolant; and, Hydriding is somewhat enhanced by neutron irradiation. Of considerable significance is the fact that a Zircaloy-4 in-reactor test section of the X-7 loop was exposed to Santowax OM at 320 to 400 o C for more than 5000 hours without excessive hydriding. (author)

Dual Microstructure Heat Treatment of a Nickel-Base Disk Alloy

Science.gov (United States)

Gayda, John

2001-01-01

Existing Dual Microstructure Heat Treat (DMHT) technology was successfully applied to Alloy 10, a high strength, nickel-base disk alloy, to produce a disk with a fine grain bore and coarse grain rim. Specimens were extracted from the DMHT disk and tested in tension, creep, fatigue, and crack growth using conditions pertinent to disk applications. These data were then compared with data from "traditional" subsolvus and supersolvus heat treatments for Alloy 10. The results showed the DMHT disk to have a high strength, fatigue resistant bore comparable to that of subsolvus Alloy 10. Further, creep resistance of the DMHT rim was comparable to that of supersolvus Alloy 10. Crack growth resistance in the DMHT rim, while better than that for subsolvus, was inferior to that of supersolvus Alloy 10. The slow cool at the end of the DMHT conversion and/or the subsolvus resolution step are thought to be responsible for degrading rim DMHT crack growth resistance.

Localized corrosion of molybdenum-bearing nickel alloys in chloride solutions

International Nuclear Information System (INIS)

Postlethwaite, J.; Scoular, R.J.; Dobbin, M.H.

1988-01-01

Electrochemical and immersion tests have been applied to a study of the localized corrosion resistance of two molybdenum-bearing nickel alloys. Alloys C-276 and 6y25, in neutral chloride solutions in the temperature range of 25 to 200 C as part of the container materials evaluation screening tests for the Canadian Nuclear Fuel Waste Management Program. Cyclic polarization studies show that the passivation breakdown potentials move rapidly to more active values with increasing temperatures, indicating a reduced resistance to localized corrosion. The results of immersion tests show that both alloys do suffer crevice corrosion in neutral aerated sodium chloride solutions at elevated temperatures, but that in both cases there is a limiting temperature > 100C, below which, the alloys are not attacked, regardless of the chloride concentration

Adhesive wear of iron chromium nickel silicon manganese molybdenum niobium alloys with duplex structure

International Nuclear Information System (INIS)

Lugscheider, E.; Deppe, E.; Ambroziak, A.; Melzer, A.

1991-01-01

Iron nickel chromium manganese silicon and iron chromium nickel manganese silicon molybdenum niobium alloys have a so-called duplex structure in a wide concentration range. This causes an excellent resistance to wear superior in the case of adhesive stress with optimized concentrations of manganese, silicon, molybdenum and niobium. The materials can be used for welded armouring structures wherever cobalt and boron-containing alloy systems are not permissible, e.g. in nuclear science. Within the framework of pre-investigations for manufacturing of filling wire electrodes, cast test pieces were set up with duplex structure, and their wear behavior was examined. (orig.) [de

Cytotoxic, allergic and genotoxic activity of a nickel-titanium alloy

NARCIS (Netherlands)

Veldhuizen, AG; Sanders, MM; Schakenraad, JM; vanHorn, [No Value

The nearly equiatomic nickel-titanium (NiTi) alloy is known for its shape memory properties. These properties can be put to excellent use in various biomedical applications, such as wires for orthodontic tooth alignment and osteosynthesis staples. The aim of this study was to evaluate the short-term

Nanophase hardfaced coatings

Energy Technology Data Exchange (ETDEWEB)

Reisgen, U.; Stein, L.; Balashov, B.; Geffers, C. [RWTH Aachen University (Germany). ISF - Welding and Joining Institute

2009-08-15

This paper demonstrates the possibility of producing iron or chromium-based nanophase hardfaced coatings by means of common arc welding methods (TIG, PTA). The appropriate composition of the alloys to be deposited allows to control the structural properties and thus also the coating properties of the weld metal. Specific variations of the alloying elements allow also the realisation of a nanostructured solidification of the carbides and borides with cooling rates that are common for arc surfacing processes. The hardfaced coatings, which had been thus produced, showed phase dimensions of approximately 100-300 nm. Based on the results it is established that the influence of the surfacing parameters and of the coating thickness and thus the influence of the heat control on the nanostructuring process is, compared with the influence of the alloy composition, of secondary importance. The generation of nanoscale structures in hardfaced coatings allows the improvement of mechanical properties, wear resistance and corrosion resistance. Potential applications for these types of hardfaced coatings lie, in particular, in the field of cutting tools that are exposed to corrosion and wear. (Abstract Copyright [2009], Wiley Periodicals, Inc.) [German] Diese Arbeit demonstriert die Moeglichkeit zur Herstellung Eisen- und Chrom-basierter nanophasiger Hartauftragschweissschichten mithilfe ueblicher Lichtbogenschweissverfahren (WIG-, Plasma-Pulver-Auftragschweissen - PPA). Eine geeignete Zusammensetzung der aufzutragenden Legierungen ermoeglicht es, die Gefuegeeigenschaften und damit die Schichteigenschaften des Schweissgutes zu kontrollieren. Gezielte Variationen der Legierungselemente erlauben die Realisierung einer nanostrukturierten Erstarrung der Karbide und Boride bei fuer Lichtbogen-Auftragschweissprozessen ueblichen Abkuehlgeschwindigkeiten. In den so erzeugten Hartschichten werden Phasengroessen von ca. 100-300 nm erreicht. Auf Basis der gewonnenen Ergebnisse kann

Precipitation hardened nickel-base alloys for sour gas environments

International Nuclear Information System (INIS)

Igarashi, M.; Mukai, S.; Kudo, T.; Okada, Y.; Ikeda, A.

1987-01-01

SCC (Stress Corrosion Cracking) in sour gas environments of γ'(gamma prime: Ni/sub 3/(Ti and/or Al)) and γ''(gamma double prime: Ni/sub 3/Nb) precipitation hardened nickel-base alloys has been studied using the SSRT (Slow Strain Rate Tensile) test, anodic polarization measurement and transmission electron microscopy (TEM). The γ'-type alloy containing Ti was more susceptible to SCC in the SSRT tests up to 350 0 F(450 K) than the γ''-type alloy containing Nb. The susceptibility to SCC was related to their deformation structures in terms of stress localization and sensitivity to pitting corrosion in H/sub 2/S solutions. TEM observation showed the γ'-type alloy deformed by the superlattice dislocations in coplanar structures. This mode of deformation induced the stress localization to some boundaries such as grain boundary and as a result the susceptibility to SCC of the γ'-type alloy was increased. On the other hand, the γ''-type alloy deformed by the massive dislocation not in coplanar structures so that it was less susceptible to SCC in terms of the stress localization. The anodic polarization measurement suggested the γ'-type alloy was more susceptible to pitting corrosion compared with the γ''-type alloy

DISSOLUTION OF ZIRCONIUM AND ALLOYS THEREFOR

Science.gov (United States)

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.

Investigations of carbon diffusion and carbide formation in nickel-based alloys

International Nuclear Information System (INIS)

Schulten, R.; Bongartz, K.; Quadakkers, W.J.; Schuster, H.; Nickel, H.

1989-11-01

The present thesis describes the carburization behaviour of nickel based alloys in heavily carburizing environments. The mechanisms of carbon diffusion and carbide precipitation in NiCr alloys with and without ternary additions of iron, cobalt or molybdenum have been investigated. Using the results of carburization experiments, a mathematical model which describes carbon diffusion and carbide formation, was developed. The simulation of the carburization process was carried out by an iterative calculation of the local thermodynamic equilibrium in the alloy. An accurate description of the carbon profiles as a function of time became possible by using a finite-difference calculation. (orig.) [de

Wear of carbide inserts with complex surface treatment when milling nickel alloy

Science.gov (United States)

Fedorov, Sergey; Swe, Min Htet; Kapitanov, Alexey; Egorov, Sergey

2018-03-01

One of the effective ways of strengthening hard alloys is the creating structure layers on their surface with the gradient distribution of physical and mechanical properties between the wear-resistant coating and the base material. The article discusses the influence of the near-surface layer which is modified by low-energy high-current electron-beam alloying and the upper anti-friction layer in a multi-component coating on the wear mechanism of the replaceable multifaceted plates in the dry milling of the difficult to machine nickel alloys.

Recycling melting process of the zirconium alloy chips

Energy Technology Data Exchange (ETDEWEB)

Reis, Luis A.M. dos; Mucsi, Cristiano S.; Tavares, Luiz A.P.; Alencar, Maicon C.; Gomes, Maurilio P.; Barbosa, Luzinete P.; Rossi, Jesualdo L., E-mail: luisreis.09@gmail.com, E-mail: csmucsi@gmail.com [Instituto de Pesquisas Energéticas e Nucleares (IPEN/CNEN-SP), São Paulo, SP (Brazil)

2017-07-01

Pressurized water reactors (PWR) commonly use {sup 235}U enriched uranium dioxide pellets as a nuclear fuel, these are assembled and stacked in zirconium alloy tubes and end caps (M5, Zirlo, Zircaloy). During the machining of these components large amounts of chips are generated which are contaminated with cutting fluid. Its storage presents safety and environmental risks due to its pyrophoric and reactive nature. Recycling industry shown interest in its recycling due to its strategic importance. This paper presents a study on the recycling process and the results aiming the efficiency in the cleaning process; the quality control; the obtaining of the pressed electrodes and finally the melting in a Vacuum Arc Remelting furnace (VAR). The recycling process begins with magnetic separation of possible ferrous alloys chips contaminant, the washing of the cutting fluid that is soluble in water, washing with an industrial degreaser, followed by a rinse with continuous flow of water under high pressure and drying with hot air. The first evaluation of the process was done by an Energy Dispersive X-rays Fluorescence Spectrometry (EDXRFS) showed the presence of 10 wt. % to 17 wt. % of impurities due the mixing with stainless steel machining chips. The chips were then pressed in a custom-made matrix of square section (40 x 40 mm - 500 mm in length), resulting in electrodes with 20% of apparent density of the original alloy. The electrode was then melted in a laboratory scale VAR furnace at the CCTM-IPEN, producing a massive ingot with 0.8 kg. It was observed that the samples obtained from Indústrias Nucleares do Brasil (INB) are supposed to be secondary scrap and it is suggested careful separation in the generation of this material. The melting of the chips is possible and feasible in a VAR furnace which reduces the storage volume by up to 40 times of this material, however, it is necessary to correct the composition of the alloy for the melting of these ingots. (author)

Recycling melting process of the zirconium alloy chips

International Nuclear Information System (INIS)

Reis, Luis A.M. dos; Mucsi, Cristiano S.; Tavares, Luiz A.P.; Alencar, Maicon C.; Gomes, Maurilio P.; Barbosa, Luzinete P.; Rossi, Jesualdo L.

2017-01-01

Pressurized water reactors (PWR) commonly use 235 U enriched uranium dioxide pellets as a nuclear fuel, these are assembled and stacked in zirconium alloy tubes and end caps (M5, Zirlo, Zircaloy). During the machining of these components large amounts of chips are generated which are contaminated with cutting fluid. Its storage presents safety and environmental risks due to its pyrophoric and reactive nature. Recycling industry shown interest in its recycling due to its strategic importance. This paper presents a study on the recycling process and the results aiming the efficiency in the cleaning process; the quality control; the obtaining of the pressed electrodes and finally the melting in a Vacuum Arc Remelting furnace (VAR). The recycling process begins with magnetic separation of possible ferrous alloys chips contaminant, the washing of the cutting fluid that is soluble in water, washing with an industrial degreaser, followed by a rinse with continuous flow of water under high pressure and drying with hot air. The first evaluation of the process was done by an Energy Dispersive X-rays Fluorescence Spectrometry (EDXRFS) showed the presence of 10 wt. % to 17 wt. % of impurities due the mixing with stainless steel machining chips. The chips were then pressed in a custom-made matrix of square section (40 x 40 mm - 500 mm in length), resulting in electrodes with 20% of apparent density of the original alloy. The electrode was then melted in a laboratory scale VAR furnace at the CCTM-IPEN, producing a massive ingot with 0.8 kg. It was observed that the samples obtained from Indústrias Nucleares do Brasil (INB) are supposed to be secondary scrap and it is suggested careful separation in the generation of this material. The melting of the chips is possible and feasible in a VAR furnace which reduces the storage volume by up to 40 times of this material, however, it is necessary to correct the composition of the alloy for the melting of these ingots. (author)

Texture Formation of Electroplated Nickel and Nickel Alloy on Cu Substrate

International Nuclear Information System (INIS)

Lee, Hee Gyoun; Hong, Gye Won; Kim, Jae Geun; Lee, Sun Wang; Kim, Ho Jin

2006-01-01

Nickel and nickel-tungsten alloy were electroplated on a cold rolled and heat treated copper(Cu) substrate. 4 mm-thick high purity commercial grade Cu was rolled to various thicknesses of 50, 70, 100 and 150 micron. High reduction ratio of 30% was applied down to 150 micron. Rolled texture was converted into cube texture via high temperature heat treatment at 400-800 degrees C. Grain size of Cu was about 50 micron which is much smaller compared to >300 micron for the Cu prepared using smaller reduction pass of 5%. 1.5 km-long 150 micron Cu was fabricated with a rolling speed of 33 m/min and texture of Cu was uniform along length. Abnormal grain growth and non-cube texture appeared for the specimen anneal above 900 degrees C. 1-10 micron thick Ni and Ni-W film was electroplated onto an annealed cube-textured Cu or directly on a cold rolled Cu. Both specimens were annealed and the degree of texture was measured. For electroplating of Ni on annealed Cu, Ni layer duplicated the cube-texture of Cu substrate and the FWHM of in plane XRD measurement for annealed Cu layer and electroplated layer was 9.9 degree and 13.4 degree, respectively. But the FWHM of in plane XRD measurement of the specimen which electroplated Ni directly on cold rolled Cu was 8.6 degree, which is better texture than that of nickel electroplated on annealed Cu and it might be caused by the suppression of secondary recrystallization and abnormal grain growth of Cu at high temperature above 900 degrees C by electroplated nickel.

Corrosion of zirconium alloys in nuclear power plants

International Nuclear Information System (INIS)

1993-01-01

To improve our understanding of corrosion mechanisms under irradiation of zirconium alloys, to collect information systematically and to identify areas where further experimentation is needed, in 1989 the IAEA initiated a special project with the participation of expert from Canada, France, Japan, USA and the former USSR. This technical document is the result of two years of joint investigations. In view of the rapidly evolving mechanistic understanding of the phenomena in this field, the document presents a series of snapshots of current ideas in specific areas of study that are relevant to the whole problem. Any attempt to present an agreed upon micromechanistic hypothesis that explains the overall phenomena must await further detailed investigations. Throughout the text, the authors have endeavored to indicate critical gaps in our basic knowledge. It is hoped that this will stimulate experimental studies in just those areas where further data are most urgently required. Refs, figs and tabs

Unirradiated UO2 in irradiated zirconium alloy sheathing

International Nuclear Information System (INIS)

MacDonald, R.D.; Hardy, D.G.; Hunt, C.E.L.; Scoberg, J.A.

1979-07-01

Zircaloy-clad UO 2 fuel elements have defected in power reactors when element power outputs were raised significantly after a long irradiation at low power. We have irradiated fuel elements fabricated from fresh UO 2 pellets and zirconium alloy sheaths previously irradiated without fuel. This gave a fuel element with radiation-damaged low-ductility sheathing but with no fission products in the fuel. The elements were power boosted in-reactor to linear power outputs up to 84 kW/m for two five-day periods. No elements defected despite sheath strains of 0.82 percent at circumferential ridge postions. Half of these elements were subsequently soaked at low power to build up the fission product inventory in the fuel and then power boosted to 63 kW/m for a third time. Two elements defected on this final boost. We conclude that these defects were caused by fission product induced stress-corrosion cracking and that this mechanism plays an importent role in power reactor fuel defects. (auth)

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

International Nuclear Information System (INIS)

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

2015-01-01

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 R dq 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

Experimental study of the zirconium alloy oxidation under high pressure of steam and modelling of the mechanisms

International Nuclear Information System (INIS)

Dali, Yacoub

2007-01-01

The corrosion of the cladding materials used for the fuel rods is one of the limiting factor of their lifetime in light water reactors. In this field, the aim of the nuclear industry is today to increase the time and the number of cycles and to submit the claddings in zirconium alloys to higher corrosive conditions. In this way, new alloys devoted to replace the standard Zircaloy-4, for instance Nb containing alloys, have been recently developed and licensed and show better corrosion resistance. A better understanding of the corrosion mechanisms of the zirconium alloys is necessary to predict the corrosion behaviour of these materials. In this work, the oxidation rate of model alloys of two metallurgic families has been studied in steam in a pressure range between 100 milli-bars and 100 bars. The Zircaloy type alloys contain as alloying elements oxygen and/or tin and/or iron and chromium. For the Zr-Nb family, three niobium contents have been studied, respectively 0.2, 0.4 and 1 weight percent of niobium. Our objectives were to understand the variations of the reactivity between the low pressure and the high pressure range, in quantifying the dependency of the corrosion rate with the steam pressure and the alloying element concentrations. The segregation process of the niobium at the surface has also been studied on the Zr-Nb alloys. During this work, a magnetic suspension thermo-balance has been developed and used to follow in-situ the corrosion rate at high pressure of water vapour. The oxide layers have been characterized by many techniques, macro and micro-photo-electrochemistry, XRD, FEG-SEM, XPS, HR-TEM and SIMS. For the Zircaloy type alloys, we have confirmed the major role of the intermetallic precipitates Zr(Fe,Cr) 2 on the corrosion resistance. Unlike the standard Zircaloy-4, for which the oxidation rate does not depend on the pressure of the water vapour and is thus limited by the vacancy diffusion in the oxide layer, we have shown that the rate of the

Straining electrode behavior and corrosion resistance of nickel base alloys in high temperature acidic solution

International Nuclear Information System (INIS)

Yamanaka, Kazuo

1992-01-01

Repassivation behavior and IGA resistance of nickel base alloys containing 0∼30 wt% chromium was investigated in high temperature acid sulfate solution. (1) The repassivation rate was increased with increasing chromium content. And so the amounts of charge caused by the metal dissolution were decreased with increasing chromium content. (2) Mill-annealed Alloy 600 suffered IGA at low pH environment below about 3.5 at the fixed potentials above the corrosion potential in 10%Na 2 SO 4 +H 2 SO 4 solution at 598K. On the other hand, thermally-treated Alloy 690 was hard to occur IGA at low pH environments which mill-annealed Alloy 600 occurred IGA. (3) It was considered that the reason, why nickel base alloys containing high chromium content such as Alloy 690 (60%Ni-30%Cr-10%Fe) had high IGA/SCC resistance in high temperature acidic solution containing sulfate ion, is due to both the promotion of the repassivation and the suppression of the film dissolution by the formation of the dense chromium oxide film

An X-ray absorption near-edge structure (XANES) study of the Sn L_3 edge in zirconium alloy oxide films formed during autoclave corrosion

International Nuclear Information System (INIS)

Hulme, Helen; Baxter, Felicity; Babu, R. Prasath; Denecke, Melissa A.; Gass, Mhairi; Steuwer, Axel; Norén, Katarina; Carlson, Stefan; Preuss, Michael

2016-01-01

Highlights: • Characterisation of tin speciation in zirconium alloy metal and oxide films using Sn L_3-XANES. • Chemical environment of tin in Zircaloy-4 and ZIRLO™ oxide films shown to be similar. • Tin in the oxide films is present in both the di- and tetravalent states and oxidises progressively with oxide-layer growth. - Abstract: Application of Sn L_3-XANES to study the oxidation state of alloying additions of tin (1–1.2 wt%) in <2 μm oxide layers formed on nuclear grade zirconium alloy has been demonstrated. Data obtained for metallic and corroded ZIRLO™ (1 wt% Sn) and Zircaloy-4 (1.2 wt% Sn) indicate tin has a similar chemical speciation in both metal alloys but this differs in the oxidised surface layers. By recording XANES at various incident angles to vary the photon penetration depth and amount of the oxide layer probed in the measurement, the authors found evidence that the oxidation of tin progresses with increasing oxide thickness.

Cast thermally stable high temperature nickel-base alloys and casting made therefrom

International Nuclear Information System (INIS)

Acuncius, D.A.; Herchenroeder, R.B.; Kirchner, R.W.; Silence, W.L.

1977-01-01

A cast thermally stable high temperature nickel-base alloy characterized by superior oxidation resistance, sustainable hot strength and retention of ductility on aging is provided by maintaining the alloy chemistry within the composition molybdenum 13.7% to 15.5%; chromium 14.7% to 16.5%; carbon up to 0.1%, lanthanum in an effective amount to provide oxidation resistance up to 0.08%; boron up to 0.015%; manganese 0.3% to 1.0%; silicon 0.2% to 0.8%; cobalt up to 2.0%; iron up to 3.0%; tungsten up to 1.0%; copper up to 0.4%; phosphorous up to 0.02%; sulfur up to 0.015%; aluminum 0.1% to 0.5% and the balance nickel while maintaining the Nv number less than 2.31

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

Science.gov (United States)

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.

Transpassive dissolution of alloy 625, chromium, nickel, and molybdenum in high-temperature solutions containing hydrochloric acid and oxygen

International Nuclear Information System (INIS)

Kritzer, P.; Boukis, N.; Dinjus, E.

2000-01-01

Coupons of nickel, molybdenum, chromium, and the nickel-based Alloy 625 (UNS 06625) were corroded in strongly oxidizing hydrochloric acid (HCl) solutions at 350 C and a pressure (p) of 24 MPa, with reaction times between 0.75 h and 50 h. For Alloy 625, the effect of surface roughness also was investigated. Nickel and molybdenum showed strong material loss after only 5 h of reaction as a result of the instability of the solid oxides formed under experimental conditions. The attack on chromium started at the grain boundaries. At longer reaction times, thick, spalling oxide layers formed on the surface. The attack on Alloy 625 also started at the grain boundaries and at inclusions leading to the formation of small pits. On polished surfaces, the growth of these pits occurred faster than on nonpolished surfaces, but fewer pits grew. Corrosion products formed at the surface consisted of oxygen and chromium. On isolated spots, nickel- and chlorine-containing products also were found

Influence of Nickel Addition on Properties of Secondary AlSi7Mg0.3 Alloy

Directory of Open Access Journals (Sweden)

Richtárech L.

2015-06-01

Full Text Available This paper deals with influence on segregation of iron based phases on the secondary alloy AlSi7Mg0.3 microstructure by nickel. Iron is the most common and harmful impurity in aluminum casting alloys and has long been associated with an increase of casting defects. In generally, iron is associated with the formation of Fe-rich intermetallic phases. It is impossible to remove iron from melt by standard operations. Some elements eliminates iron by changing iron intermetallic phase morphology, decreasing its extent and by improving alloy properties. Realization of experiments and results of analysis show new view on solubility of iron based phases during melt preparation with higher iron content and influence of nickel as iron corrector of iron based phases.

Electrochemical evaluation of zinc effect on the corrosion of nickel alloy in PWR solutions with increasing temperature

International Nuclear Information System (INIS)

Alvial M, Gaston; Neves, Celia F.C.; Schvartzman, Monica M.A.M.; Quinan, Marco Antonio D.

2007-01-01

The main objective for the addition of zinc acetate to the reactor coolant system of PWRs is to effect radiation dose rate reductions. However, zinc is also added as an approach to mitigate the occurrence or severity of primary water stress corrosion cracking of nickel alloy 600. The mechanism by which zinc affects the corrosion of austenitic nickel-base alloys is by incorporation of zinc into the spinel oxide corrosion films. The purpose of this work is to evaluate the influence of zinc on the corrosion behavior of the nickel alloy 600 in PWR chemical environment (1200 ppm B, 2.2 ppm Li, deoxygenated water) with increasing temperature at room pressure. Electrochemical tests (anodic potentiodynamic polarization and electrochemical impedance spectroscopy) were used to characterize the alloy 600. Two conditions were applied: 0 and 100 ppb zinc and the temperature range was 50 - 90 deg C, at ambient pressure. Potentiodynamic polarization was inefficient to present conclusive results. Impedance measurements showed single semicircle in the Nyquist plane suggesting reduction of the charge transference resistance in zinc-containing solutions. This effect is evident at 90 deg C suggesting prejudicial influence of zinc for the alloy 600 at room pressure. (author)

Alloy spreading and filling of gaps in brazing of VDU-2 and KhN50VMTYuB heat resistant nickel alloys with VPr3K and VPr10 alloys

International Nuclear Information System (INIS)

Shapiro, A.E.; Podol'skij, B.A.; Lepisko, M.R.; Borzyak, A.G.; Moryakov, V.F.; Rostislavskaya, T.T.

1984-01-01

A study was made on contact interaction of VDU-2 and KhN50VMTYuB alloys with VPr3K and VPr10 alloys at 1325 and 1220 deg C in argon and industrial vacuum. The contact angles and wettability indexes were determined. The solders fill the vertical gaps of up to 0.25 mm width through 80 mm height. Spreading and filling of gaps proceeds better during soldering in argon with boron trifluoride addition as compared to soldering in industrial vacuum. VPr10 alloy is divided into two phases when wetting KhN50VMTYuB alloy: fusible one on the base of nickel-chromium-manganese solution and infusible one on the base of nickel-niobium eutectics. The square of fusible phase spreading is 2.5...3 times larger as compared to infusible one

Stress corrosion cracking of nickel alloys in bicarbonate and chloride solutions

International Nuclear Information System (INIS)

Ares, A. E.; Carranza, R. M.; Giordano, C. M.; Zadorozne, N. S.; Rebak, R.B.

2013-01-01

Alloy 22 is one of the candidates for the manufacture of high level radioactive waste containers. These containers provide services in natural environments characterized by multi-ionics solutions, it is estimated they could suffer three types of deterioration: general corrosion, localized corrosion (crevice corrosion) and stress corrosion cracking (SCC). It has been confirmed that the presence of bicarbonate at temperatures above 60°C and applied potentials around +400 mVSCE are necessary in order to produce cracking, . This susceptibility may be associated to the instability of the passive film formed and to the formation of an anodic current peak in the polarization curves in these media. Until now, it is unclear the role played by each alloying element (Ni, Cr or Mo) in the SCC susceptibility of Alloy 22 in these media The aim of this work is to evaluate the SCC susceptibility of nickel-based alloys in media containing bicarbonate and chloride ions, at high temperature. Slow Strain Rate Testing (SSRT) was conducted to samples of different alloys: 22 (Ni-Cr-Mo), 600 (Ni-Cr-Fe), 800H (Ni-Fe-Cr) y 201 (99.5% Ni).This tests were conducted in 1.1 mol/L NaHCO 3 +1.5 mol/L NaCl a 90°C and different applied potentials (+200mVSCE,+300 mVSCE, +400 mVSCE). These results were complemented with those obtained in a previous work, where we studied the anodic electrochemical behavior of nickel base alloys under the same conditions. It was found that alloy 22 showed a current peak in a potential range between +200 mVSCE and +300 mVSCE when immersed in bicarbonate ions containing solutions. This peak was attributed to the presence of chromium in the alloys. The SSRT showed that only alloy 22 has a clear indication of stress corrosion cracking. The current results suggested that the presence of an anodic peak in the polarization curves was not a sufficient condition for cracking. (author)

A new model for prediction of dispersoid precipitation in aluminium alloys containing zirconium and scandium

International Nuclear Information System (INIS)

Robson, J.D.

2004-01-01

A model has been developed to predict precipitation of ternary Al 3 (Sc, Zr) dispersoids in aluminium alloys containing zirconium and scandium. The model is based on the classical numerical method of Kampmann and Wagner, extended to predict precipitation of a ternary phase. The model has been applied to the precipitation of dispersoids in scandium containing AA7050. The dispersoid precipitation kinetics and number density are predicted to be sensitive to the scandium concentration, whilst the dispersoid radius is not. The dispersoids are predicted to enrich in zirconium during precipitation. Coarsening has been investigated in detail and it has been predicted that a steady-state size distribution is only reached once coarsening is well advanced. The addition of scandium is predicted to eliminate the dispersoid free zones observed in scandium free 7050, greatly increasing recrystallization resistance

Standard specification for cobalt-chromium-nickel-molybdenum-tungsten alloy (UNS R31233) plate, sheet and strip. ASTM standard

International Nuclear Information System (INIS)

1998-09-01

This specification is under the jurisdiction of ASTM Committee B-2 on Nonferrous Metals and Alloys and is the direct responsibility of Subcommittee B02.07 on Refined Nickel and Cobalt, and Alloys Containing Nickel or Cobalt or Both as Principal Constituents. Current edition approved Apr. 10, 1998 and published September 1998. Originally published as B 818-91. Last previous edition was B 818-93

Synthesis and Tribological Performance of Different Particle-Sized Nickel-Ion-Exchanged α-Zirconium Phosphates

Science.gov (United States)

Zhang, Xiaosheng; Xu, Hong; Dong, Jinxiang

2018-03-01

Nickel-ion-exchanged α-zirconium phosphate (Ni-α-ZrP) was synthesized by a mild hydrothermal synthesis method. Different raw material ratios (NaF/H3PO4/Ni(CH3COO)2·4H2O) influence the particle size of the Ni-α-ZrP samples. The grain size could be controlled and distributed from 20 to 600 nm. Ni-α-ZrP was evaluated as an additive in lithium grease in a four-ball test. A 3.0 wt.% addition of Ni-α-ZrP to lithium grease yielded maximum non-seizure load values of 1235 N, and the wear scar diameter on the lower balls is 0.42 mm at 294 N. Compared with smaller particles, the addition of Ni-α-ZrP with a larger particle size to grease yields a better load-carrying capacity.

Research into zirconium alloys resistant to carbon dioxide under pressure at temperatures of up to 600 deg C (1963)

International Nuclear Information System (INIS)

Baque, P.; Dominget, R.; Bossard, J.

1963-01-01

Zirconium is a metal having a relatively low neutron capture cross-section and a high melting point; it is thus possible to consider its use in particular as a canning material for fuel elements in CO 2 -cooled nuclear reactors. A preliminary study of several types of zirconium showed that the metal is already strongly oxidised in this gas at 500 deg C. The 'breakaway' phenomenon is generalised; the oxidation rate is then linear and depends on the carbon dioxide pressure. An attempt was therefore made to find binary and tertiary alloys in order to improve the metal behaviour. Several interesting compositions were found: 1, 1.6 and 2.5 per cent of copper, 2 per cent of vanadium, and 0.05 and 0.5 per cent of calcium. Tertiary copper-molybdenum and copper-phosphorus alloys are also less liable to oxidation and in particular do not exhibit the 'breakaway' phenomenon even after a prolonged treatment at 600 deg C. (authors) [fr

Polarographic determination of the titanium and niobium content of zirconium alloys

International Nuclear Information System (INIS)

Levin, R; Gabra, J.

1978-03-01

A method is described for the polarographic determination of titanium and niobium in zirconium alloys in the concentration range of 0.1% to 4% of each of the determined metals. To assure the complete dissolution of the sample a mixture of nitric acid and hydrofluoric acid is used. After evaporating these acids in the presence of sulphuric acid, the contents are determined polarographically with a supporting electrolyte solution of 0.1M EDTA, 0.33M potassium sulfate and 0.4M sodium acetate, buffered to pH 4 with acetic acid. The half-wave potential (Esub(1/2)) of titanium is -0.35V and that of niobium is -0.67 V. (author)

Thermogravimetric study of reduction of oxides present in oxidized nickel-base alloy powders

Science.gov (United States)

Herbell, T. P.

1976-01-01

Carbon, hydrogen, and hydrogen plus carbon reduction of three oxidized nickel-base alloy powders (a solid solution strengthened alloy both with and without the gamma prime formers aluminum and titanium and the solid solution strengthened alloy NiCrAlY) were evaluated by thermogravimetry. Hydrogen and hydrogen plus carbon were completely effective in reducing an alloy containing chromium, columbium, tantalum, molybdenum, and tungsten. However, with aluminum and titanium present the reduction was limited to a weight loss of about 81 percent. Carbon alone was not effective in reducing any of the alloys, and none of the reducing conditions were effective for use with NiCrAlY.

Ostwald ripening of Pb nanocrystalline phase in mechanically milled Al-Pb alloys and the influence of Cu additive

International Nuclear Information System (INIS)

Wu, Z.F.; Zeng, M.Q.; Ouyang, L.Z.; Zhang, X.P.; Zhu, M.

2005-01-01

The coarsening behavior of nanosized Pb phase in both Al-10%Pb and Al-10%Pb-4.5%Cu alloys has been studied by X-ray diffraction and transmission electron microscopy analysis. The coarsening of Pb nanophase in Al-Pb alloys still follows the classical ripening theory (the LSW theory) and the addition of Cu decreases the coarsening rate of Pb nanophase

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

Science.gov (United States)

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.

Structure determination of electrodeposited zinc-nickel alloys: thermal stability and quantification using XRD and potentiodynamic dissolution

International Nuclear Information System (INIS)

Fedi, B.; Gigandet, M.P.; Hihn, J-Y; Mierzejewski, S.

2016-01-01

Highlights: • Quantification of zinc-nickel phases between 1,2% and 20%. • Coupling XRD to partial potentiodynamic dissolution. • Deconvolution of anodic stripping curves. • Phase quantification after annealing. - Abstract: Electrodeposited zinc-nickel coatings obtained by electrodeposition reveal the presence of metastable phases in various quantities, thus requiring their identification, a study of their thermal stability, and, finally, determination of their respective proportions. By combining XRD measurement with partial potentiodynamic dissolution, anodic peaks were indexed to allow their quantification. Quantification of electrodeposited zinc-nickel alloys approximately 10 μm thick was thus carried out on nickel content between 1.2% and 20%, and exhibited good accuracy. This method was then extended to the same set of alloys after annealing (250 °C, 2 h), thus bringing the structural organization closer to its thermodynamic equilibrium. The result obtained ensures better understanding of crystallization of metastable phases and of phase proportion evolution in a bi-phasic zinc-nickel coating. Finally, the presence of a monophase γ and its thermal stability in the 12% to 15% range provides important information for coating anti-corrosion behavior.

The technologies of zirconium production for nuclear fuel components in Ukraine

International Nuclear Information System (INIS)

Semenov, G.R.

2000-01-01

Perspectives of development zirconium alloys and WWER-1000 assemble components production in Ukraine are considered. Basic technological production processes of zirconium alloys in conditions of Ukrainian enterprises and modern requirements are analyzed. The critical processes on technical and economic criteria are defined. The main directions of activity and steps on technological processes improvement for production quality providing are offered. (author)

Nickel: makes stainless steel strong

Science.gov (United States)

Boland, Maeve A.

2012-01-01

Nickel is a silvery-white metal that is used mainly to make stainless steel and other alloys stronger and better able to withstand extreme temperatures and corrosive environments. Nickel was first identified as a unique element in 1751 by Baron Axel Fredrik Cronstedt, a Swedish mineralogist and chemist. He originally called the element kupfernickel because it was found in rock that looked like copper (kupfer) ore and because miners thought that "bad spirits" (nickel) in the rock were making it difficult for them to extract copper from it. Approximately 80 percent of the primary (not recycled) nickel consumed in the United States in 2011 was used in alloys, such as stainless steel and superalloys. Because nickel increases an alloy's resistance to corrosion and its ability to withstand extreme temperatures, equipment and parts made of nickel-bearing alloys are often used in harsh environments, such as those in chemical plants, petroleum refineries, jet engines, power generation facilities, and offshore installations. Medical equipment, cookware, and cutlery are often made of stainless steel because it is easy to clean and sterilize. All U.S. circulating coins except the penny are made of alloys that contain nickel. Nickel alloys are increasingly being used in making rechargeable batteries for portable computers, power tools, and hybrid and electric vehicles. Nickel is also plated onto such items as bathroom fixtures to reduce corrosion and provide an attractive finish.

The strengthening mechanism of a nickel-based alloy after laser shock processing at high temperatures

International Nuclear Information System (INIS)

Li, Yinghong; Zhou, Liucheng; He, Weifeng; He, Guangyu; Wang, Xuede; Nie, Xiangfan; Wang, Bo; Luo, Sihai; Li, Yuqin

2013-01-01

We investigated the strengthening mechanism of laser shock processing (LSP) at high temperatures in the K417 nickel-based alloy. Using a laser-induced shock wave, residual compressive stresses and nanocrystals with a length of 30–200 nm and a thickness of 1 μm are produced on the surface of the nickel-based alloy K417. When the K417 alloy is subjected to heat treatment at 900 °C after LSP, most of the residual compressive stress relaxes while the microhardness retains good thermal stability; the nanocrystalline surface has not obviously grown after the 900 °C per 10 h heat treatment, which shows a comparatively good thermal stability. There are several reasons for the good thermal stability of the nanocrystalline surface, such as the low value of cold hardening of LSP, extreme high-density defects and the grain boundary pinning of an impure element. The results of the vibration fatigue experiments show that the fatigue strength of K417 alloy is enhanced and improved from 110 to 285 MPa after LSP. After the 900 °C per 10 h heat treatment, the fatigue strength is 225 MPa; the heat treatment has not significantly reduced the reinforcement effect. The feature of the LSP strengthening mechanism of nickel-based alloy at a high temperature is the co-working effect of the nanocrystalline surface and the residual compressive stress after thermal relaxation. (paper)

Orientation dependence of the thermal fatigue of nickel alloy single crystals

Energy Technology Data Exchange (ETDEWEB)

Dul' nev, R A; Svetlov, I L; Bychkov, N G; Rybina, T V; Sukhanov, N N

1988-11-01

The orientation dependence of the thermal stability and the thermal fatigue fracture characteristics of single crystals of MAR-M200 nickel alloy are investigated experimentally using X-ray diffraction analysis and optical and scanning electron microscopy. It is found that specimens with the 111-line orientation have the highest thermal stability and fatigue strength. Under similar test conditions, the thermal fatigue life of single crystals is shown to be a factor of 1.5-2 higher than that of the directionally solidified and equiaxed alloys. 6 references.

Investigation of hydrogen evolution activity for the nickel, nickel-molybdenum nickel-graphite composite and nickel-reduced graphene oxide composite coatings

International Nuclear Information System (INIS)

Jinlong, Lv; Tongxiang, Liang; Chen, Wang

2016-01-01

Graphical abstract: - Highlights: • Improved HER efficiency of Ni-Mo coatings was attributed to ‘cauliflower’ like microstructure. • RGO in nickel-RGO composite coating promoted refined grain and facilitated HER. • Synergistic effect between nickel and RGO facilitated HER due to large specific surface of RGO. - Abstract: The nickel, nickel-molybdenum alloy, nickel-graphite and nickel-reduced graphene oxide composite coatings were obtained by the electrodeposition technique from a nickel sulfate bath. Nanocrystalline molybdenum, graphite and reduced graphene oxide in nickel coatings promoted hydrogen evolution reaction in 0.5 M H_2SO_4 solution at room temperature. However, the nickel-reduced graphene oxide composite coating exhibited the highest electrocatalytic activity for the hydrogen evolution reaction in 0.5 M H_2SO_4 solution at room temperature. A large number of gaps between ‘cauliflower’ like grains could decrease effective area for hydrogen evolution reaction in slight amorphous nickel-molybdenum alloy. The synergistic effect between nickel and reduced graphene oxide promoted hydrogen evolution, moreover, refined grain in nickel-reduced graphene oxide composite coating and large specific surface of reduced graphene oxide also facilitated hydrogen evolution reaction.

Investigation of hydrogen evolution activity for the nickel, nickel-molybdenum nickel-graphite composite and nickel-reduced graphene oxide composite coatings

Energy Technology Data Exchange (ETDEWEB)

Jinlong, Lv, E-mail: ljlbuaa@126.com [Beijing Key Laboratory of Fine Ceramics, Institute of Nuclear and New Energy Technology, Tsinghua University, Zhongguancun Street, Haidian District, Beijing 100084 (China); State Key Lab of New Ceramic and Fine Processing, Tsinghua University, Beijing 100084 (China); Tongxiang, Liang; Chen, Wang [Beijing Key Laboratory of Fine Ceramics, Institute of Nuclear and New Energy Technology, Tsinghua University, Zhongguancun Street, Haidian District, Beijing 100084 (China); State Key Lab of New Ceramic and Fine Processing, Tsinghua University, Beijing 100084 (China)

2016-03-15

Graphical abstract: - Highlights: • Improved HER efficiency of Ni-Mo coatings was attributed to ‘cauliflower’ like microstructure. • RGO in nickel-RGO composite coating promoted refined grain and facilitated HER. • Synergistic effect between nickel and RGO facilitated HER due to large specific surface of RGO. - Abstract: The nickel, nickel-molybdenum alloy, nickel-graphite and nickel-reduced graphene oxide composite coatings were obtained by the electrodeposition technique from a nickel sulfate bath. Nanocrystalline molybdenum, graphite and reduced graphene oxide in nickel coatings promoted hydrogen evolution reaction in 0.5 M H{sub 2}SO{sub 4} solution at room temperature. However, the nickel-reduced graphene oxide composite coating exhibited the highest electrocatalytic activity for the hydrogen evolution reaction in 0.5 M H{sub 2}SO{sub 4} solution at room temperature. A large number of gaps between ‘cauliflower’ like grains could decrease effective area for hydrogen evolution reaction in slight amorphous nickel-molybdenum alloy. The synergistic effect between nickel and reduced graphene oxide promoted hydrogen evolution, moreover, refined grain in nickel-reduced graphene oxide composite coating and large specific surface of reduced graphene oxide also facilitated hydrogen evolution reaction.

The Hydrogen Pickup Behavior for Zirconium-based Alloys in Various Out-of-pile Corrosion Test Conditions

Energy Technology Data Exchange (ETDEWEB)

Aomi, M.; Etoh, Y.; Ishimoto, S.; Une, K. [Nippon Nuclear Fuel Development, Co., Ltd., 2163 Narita-cho, Oarai-machi, Ibaraki-ken, 311-1313 (Japan); Ito, K. [Global Nuclear Fuel Japan Co., Ltd., 3-1 Uchikawa 2-chome, Yokosuka-shi, Kanagawa-ken, 239-0836 (Japan)

2009-06-15

An acceleration of hydrogen absorption in zirconium alloy claddings at high burnups is one of the most important issues limiting the fuel performance from the viewpoint of cladding integrity. In this context, advanced cladding materials with higher corrosion resistant and lower hydrogen absorption properties have been widely searched in various organizations. In this study, four kinds of zirconium-based alloys, whose in-pile data had been acquired [1,2] were subjected to comprehensive out-of-pile corrosion tests with various temperature and atmosphere conditions in order to investigate the correlation between in-pile and out-of-pile corrosion and hydrogen pick-up behavior, i.e. Zry-2, GNF-Ziron (Zry-2-based alloy with {approx}0.25 wt % of Fe), Hi-FeNi Zircaloy (Zry-2-based alloy with {approx}0.25 wt % of Fe and {approx}0.1 wt% Ni), and VB (Zr-based alloy containing Sn, Cr, and {approx}0.5 wt % of Fe). All the alloys were annealed in RXA condition. The out-of-pile corrosion tests were carried out in three different conditions of 400 deg. C steam, 475 deg. C supercritical water, and 290 deg. C LiOH aqueous solution. In addition to these alloys, several Zry-2-based alloys with various iron contents were tested in 290 deg. C LiOH aqueous solution. Among the four corrosion conditions, the 290 deg. C LiOH aqueous solution test well screened the hydrogen pick-up behavior of the alloys. The hydrogen absorption decreased with higher iron contents in the alloys in both the out-of-pile and in-pile conditions. Especially, the distinct suppression of hydrogen absorption was observed for VB with the highest iron content. The similar dependence of iron content on the hydrogen pick-up fraction was also obtained for the Zry-2-based alloys with different iron contents, which were corroded in the 290 deg. C LiOH aqueous solution condition. As for the corrosion behavior in the 290 deg. C LiOH aqueous solution condition, the weight gains of Zry-2, GNF-Ziron and VB followed the 1

On the initial corrosion mechanism of zirconium alloy: Interaction of oxygen and water with Zircaloy at room temperature and 450 C evaluated by x-ray absorption spectroscopy and photoelectron spectroscopy

International Nuclear Information System (INIS)

Doebler, U.; Knop, A.

1994-01-01

The initial stages of zirconium oxide formation on Zircaloy after water (H 2 O) and oxygen (O 2 ) exposures have been investigated in situ using photoelectron spectroscopy and X-ray-absorption spectroscopy. The reactivity of the zirconium alloy with O 2 at room temperature is about 1,000 times higher than for H 2 O. Up to 100 L (1 L = 1 Langmuir unit = 1 · 10 -6 mbar · s) H 2 O exposure, the reactivity of the zirconium alloy at 450 C is comparable to the room temperature reaction. At higher H 2 O exposure, a sharp increase in the reaction rate for the high-temperature oxidation is observed. From the energy position of the Zr 3d photo emission line and their oxygen-induced chemical shifts, one can really follow the formation of the oxide films. Two different substoichiometric oxides were found during reaction with water. Suboxide (1) is located at the zirconium/zirconium-oxide interface. Subsequently, a Suboxide (2) is concluded from the chemical shift of the zirconium photoelectrons. After an oxide thickness of 2 nm, the stoichiometric ZrO 2 phase is not yet developed

Calculation of the driving force for the radiation induced precipitation of Ni3Si in nickel-silicon alloys

International Nuclear Information System (INIS)

Miodownik, A.P.; Watkin, J.S.

1979-01-01

The appearance of precipitates which have been identified as Ni 3 Si in irradiated stainless steels and nickel rich alloys such as Inconel is of considerable interest in relation to the swelling behaviour of such materials. Work on binary nickel-silicon alloys has shown that Ni 3 Si can be induced to precipitate in alloys whose silicon content is well below the accepted solubility limit, and it has also been shown that such precipitates redissolve when heat-treatment is continued at the same temperature in the absence of irradiation. Such effects imply an irradiation induced shift of chemical potential, and cannot be explained by merely involving accelerated diffusion. This paper represents an attempt to calculate the shift in chemical potential required to precipitate Ni 3 Si in alloys containing 1-10% Si (at%) over a range of temperatures (300-1000K), and then proceeds to relate this calculated chemical potential with available information concerning the dose rates required to induce such precipitates at various temperatures. Presentation of the results is modelled on the well established methods for handling the Time-Temperature-Transformation behaviour of ordinary alloy systems, with dose rate being substituted for the time axis. Analogous calculations are presented for nickel-germanium alloys, in order to check whether the numerical values deduced from the nickel silicon system have more general applicability, and also to see whether there are any significant differences in a system where the size factor of the solute is of the opposite sign. (orig.) [de

Generalized corrosion of nickel base alloys in high temperature aqueous media: a contribution to the comprehension of the mechanisms

International Nuclear Information System (INIS)

Marchetti-Sillans, L.

2007-11-01

In France, nickel base alloys, such as alloy 600 and alloy 690, are the materials constituting steam generators (SG) tubes of pressurized water reactors (PWR). The generalized corrosion resulting from the interaction between these alloys and the PWR primary media leads, on the one hand, to the formation of a thin protective oxide scale (∼ 10 nm), and on the other hand, to the release of cations in the primary circuit, which entails an increase of the global radioactivity of this circuit. The goal of this work is to supply some new comprehension elements about nickel base alloys corrosion phenomena in PWR primary media, taking up with underlining the effects of metallurgical and physico-chemical parameters on the nature and the growth mechanisms of the protective oxide scale. In this context, the passive film formed during the exposition of alloys 600, 690 and Ni-30Cr, in conditions simulating the PWR primary media, has been analyzed by a set of characterization techniques (SEM, TEM, PEC and MPEC, XPS). The coupling of these methods leads to a fine description, in terms of nature and structure, of the multilayered oxide forming during the exposition of nickel base alloys in primary media. Thus, the protective part of the oxide scale is composed of a continuous layer of iron and nickel mixed chromite, and Cr 2 O 3 nodules dispersed at the alloy / mixed chromite interface. The study of protective scale growth mechanisms by tracers and markers experiments reveals that the formation of the mixed chromite is the consequence of an anionic mechanism, resulting from short circuits like grain boundaries diffusion. Besides, the impact of alloy surface defects has also been studied, underlining a double effect of this parameter, which influences the short circuits diffusion density in oxide and the formation rate of Cr 2 O 3 nodules. The sum of these results leads to suggest a description of the nickel base alloys corrosion mechanisms in PWR primary media and to tackle some

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

Science.gov (United States)

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.

Optical modeling of nickel-base alloys oxidized in pressurized water reactor

Energy Technology Data Exchange (ETDEWEB)

Clair, A. [Laboratoire Interdisciplinaire Carnot de Bourgogne, UMR 6303 CNRS, Universite de Bourgogne, 9 avenue Alain Savary, BP 47870, 21078 Dijon cedex (France); Foucault, M.; Calonne, O. [Areva ANP, Centre Technique Departement Corrosion-Chimie, 30 Bd de l' industrie, BP 181, 71205 Le Creusot (France); Finot, E., E-mail: Eric.Finot@u-bourgogne.fr [Laboratoire Interdisciplinaire Carnot de Bourgogne, UMR 6303 CNRS, Universite de Bourgogne, 9 avenue Alain Savary, BP 47870, 21078 Dijon cedex (France)

2012-10-01

The knowledge of the aging process involved in the primary water of pressurized water reactor entails investigating a mixed growth mechanism in the corrosion of nickel-base alloys. A mixed growth induces an anionic inner oxide and a cationic diffusion parallel to a dissolution-precipitation process forms the outer zone. The in situ monitoring of the oxidation kinetics requires the modeling of the oxide layer stratification with the full knowledge of the optical constants related to each component. Here, we report the dielectric constants of the alloys 600 and 690 measured by spectroscopic ellipsometry and fitted to a Drude-Lorentz model. A robust optical stratification model was determined using focused ion beam cross-section of thin foils examined by transmission electron microscopy. Dielectric constants of the inner oxide layer depleted in chromium were assimilated to those of the nickel thin film. The optical constants of both the spinels and extern layer were determined. - Highlights: Black-Right-Pointing-Pointer Spectroscopic ellipsometry of Ni-base alloy oxidation in pressurized water reactor Black-Right-Pointing-Pointer Measurements of the dielectric constants of the alloys Black-Right-Pointing-Pointer Optical simulation of the mixed oxidation process using a three stack model Black-Right-Pointing-Pointer Scattered crystallites cationic outer layer; linear Ni-gradient bottom layer Black-Right-Pointing-Pointer Determination of the refractive index of the spinel and the Cr{sub 2}O{sub 3} layers.

Contribution to the study of the electrodeposition of iron-nickel alloys; Contribution a l'etude du depot electrolytique des alliages fer-nickel

Energy Technology Data Exchange (ETDEWEB)

Valignat, J [Commissariat a l' Energie Atomique, Grenoble (France). Centre d' Etudes Nucleaires

1967-12-01

Using a coulometric technique based upon the anodic intentiostatic dissolution, we studied the potentiostatic, deposition of nickel, iron and nickel iron alloys. We have shown that the minimum of the curve I = f (t) (deposition current versus time) is probably due to the transitory blocking of the surface by hydrogen and that the syn-crystallisation of nickel and iron is responsible for the anomalous co-deposition of these two elements. (author) [French] En employant une methode coulometrique par dissolution anodique intensipstatique, nous avons etudie le depot potentiostatique du nickel, du fer et des alliages fer-nickel. Nous avons pu montrer que le minimum de la courbe I = f (t) enregistree au cours du depot est du probablement au blocage momentane de la surface par l'hydrogene et que la syncristallisation du fer et du nickel est responsable de l'anomalie du depot simultane de ces deux elements. (auteur)

Sonocatalytic injury of cancer cells attached on the surface of a nickel-titanium dioxide alloy plate.

Science.gov (United States)

Ninomiya, Kazuaki; Maruyama, Hirotaka; Ogino, Chiaki; Takahashi, Kenji; Shimizu, Nobuaki

2016-01-01

The present study demonstrates ultrasound-induced cell injury using a nickel-titanium dioxide (Ni-TiO2) alloy plate as a sonocatalyst and a cell culture surface. Ultrasound irradiation of cell-free Ni-TiO2 alloy plates with 1 MHz ultrasound at 0.5 W/cm(2) for 30s led to an increased generation of hydroxyl (OH) radicals compared to nickel-titanium (Ni-Ti) control alloy plates with and without ultrasound irradiation. When human breast cancer cells (MCF-7 cells) cultured on the Ni-TiO2 alloy plates were irradiated with 1 MHz ultrasound at 0.5 W/cm(2) for 30s and then incubated for 48 h, cell density on the alloy plate was reduced to approximately 50% of the controls on the Ni-Ti alloy plates with and without ultrasound irradiation. These results indicate the injury of MCF-7 cells following sonocatalytic OH radical generation by Ni-TiO2. Further experiments demonstrated cell shrinkage and chromatin condensation after ultrasound irradiation of MCF-7 cells attached on the Ni-TiO2 alloy plates, indicating induction of apoptosis. Copyright © 2015 Elsevier B.V. All rights reserved.

Experimental approach and micro-mechanical modeling of the mechanical behavior of irradiated zirconium alloys

International Nuclear Information System (INIS)

Onimus, F.

2003-12-01

Zirconium alloys cladding tubes containing nuclear fuel of the Pressurized Water Reactors constitute the first safety barrier against the dissemination of radioactive elements. Thus, it is essential to predict the mechanical behavior of the material in-reactor conditions. This study aims, on the one hand, to identify and characterize the mechanisms of the plastic deformation of irradiated zirconium alloys and, on the other hand, to propose a micro-mechanical modeling based on these mechanisms. The experimental analysis shows that, for the irradiated material, the plastic deformation occurs by dislocation channeling. For transverse tensile test and internal pressure test this channeling occurs in the basal planes. However, for axial tensile test, the study revealed that the plastic deformation also occurs by channeling but in the prismatic and pyramidal planes. In addition, the study of the macroscopic mechanical behavior, compared to the deformation mechanisms observed by TEM, suggested that the internal stress is higher in the case of irradiated material than in the case of non-irradiated material, because of the very heterogeneous character of the plastic deformation. This analysis led to a coherent interpretation of the mechanical behavior of irradiated materials, in terms of deformation mechanisms. The mechanical behavior of irradiated materials was finally modeled by applying homogenization methods for heterogeneous materials. This model is able to reproduce adequately the mechanical behavior of the irradiated material, in agreement with the TEM observations. (author)

The Influence of Nickel and Tin Additives on the Microstructural and Mechanical Properties of Al-Zn-Mg-Cu Alloys

Directory of Open Access Journals (Sweden)

Haider T. Naeem

2014-01-01

Full Text Available The effects of nickel and nickel combined tin additions on mechanical properties and microstructural evolutions of aluminum-zinc-magnesium-copper alloys were investigated. Aluminum alloys containing Ni and Sn additives were homogenized at different temperatures conditions and then aged at 120°C for 24 h (T6 and retrogressed at 180°C for 30 min and then reaged at 120°C for 24 h (RRA. Comparison of the ultimate tensile strength (UTS of as-quenched Al-Zn-Mg-Cu-Ni and Al-Zn-Mg-Cu-Ni-Sn alloys with that of similar alloys which underwent aging treatment at T6 temper showed that gains in tensile strengths by 385 MPa and 370 MPa were attained, respectively. These improvements are attributed to the precipitation hardening effects of the alloying element within the base alloy and the formation of nickel/tin-rich dispersoid compounds. These intermetallic compounds retard the grain growth, lead to grain refinement, and result in further strengthening effects. The outcomes of the retrogression and reaging processes which were carried on aluminum alloys indicate that the mechanical strength and Vickers hardness have been enhanced much better than under the aging at T6 temper.

Oxidation of zirconium-aluminum alloys

International Nuclear Information System (INIS)

Cox, B.

1967-10-01

Examination of the processes occurring during the oxidation of Zr-1% A1, Zr-3% A1, and Zr-1.5% A1-0.5% Mo alloys has shown that in steam rapid oxidation occurs predominantly around the Zr 3 A1 particles, which at low temperatures appear to be relatively unattacked. The unoxidised particles become incorporated in the oxide, and become fully oxidised as the film thickens. This rapid localised oxidation is preceded by a short period of uniform film growth, during which the oxide film thickness does not exceed ∼200A-o. Thus the high oxidation rates can probably be ascribed to aluminum in solution in the zirconium matrix, although its precise mode of operation has not been determined. Once the solubility limit of aluminum is exceeded, the size, distribution and number of intermetallic particles affects the oxidation rate merely by altering the distribution of regions of metal giving high oxidation rates. The controlling process during the early stages of oxidation is electron transport and not ionic transport. Thus, the aluminum in the oxide film is presumably increasing the ionic conductivity more than the electronic. The oxidation rates in atmospheric pressure steam are very high and their irregular temperature dependence suggests that the oxidation rate will be pressure dependent. This was confirmed, in part, by a comparison with oxidation in moist air. It was found that the rate of development of white oxide around intermetallic particles was considerably reduced by the decrease in the partial pressure of H 2 O; the incubation period was not much different, however. (author)

Modeling of Some Physical Properties of Zirconium Alloys for Nuclear Applications in Support of UFD Campaign

Energy Technology Data Exchange (ETDEWEB)

Michael V. Glazoff

2013-08-01

Zirconium-based alloys Zircaloy-2 and Zircaloy-4 are widely used in the nuclear industry as cladding materials for light water reactor (LWR) fuels. These materials display a very good combination of properties such as low neutron absorption, creep behavior, stress-corrosion cracking resistance, reduced hydrogen uptake, corrosion and/or oxidation, especially in the case of Zircaloy-4. However, over the last couple of years, in the post-Fukushima Daiichi world, energetic efforts have been undertaken to improve fuel clad oxidation resistance during off-normal temperature excursions. Efforts have also been made to improve upon the already achieved levels of mechanical behavior and reduce hydrogen uptake. In order to facilitate the development of such novel materials, it is very important to achieve not only engineering control, but also a scientific understanding of the underlying material degradation mechanisms, both in working conditions and in storage of used nuclear fuel. This report strives to contribute to these efforts by constructing the thermodynamic models of both alloys; constructing of the respective phase diagrams, and oxidation mechanisms. A special emphasis was placed upon the role of zirconium suboxides in hydrogen uptake reduction and the atomic mechanisms of oxidation. To that end, computational thermodynamics calculations were conducted concurrently with first-principles atomistic modeling.

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

Energy Technology Data Exchange (ETDEWEB)

Platt, P., E-mail: Philip.Platt@manchester.ac.uk [University of Manchester, School of Materials, Materials Performance Centre, Manchester M13 9PL (United Kingdom); Frankel, P. [University of Manchester, School of Materials, Materials Performance Centre, Manchester M13 9PL (United Kingdom); Gass, M.; Howells, R. [AMEC, Walton House, Faraday Street, Birchwood Park, Risley, Warrington WA3 6GA (United Kingdom); Preuss, M. [University of Manchester, School of Materials, Materials Performance Centre, Manchester M13 9PL (United Kingdom)

2014-11-15

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.

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

Science.gov (United States)

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.

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

International Nuclear Information System (INIS)

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

2014-01-01

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

Development of phased array UT procedure for crack depth sizing on nickel based alloy weld

International Nuclear Information System (INIS)

Hirasawa, Taiji; Okada, Hisao; Fukutomi, Hiroyuki

2012-01-01

Recently, it is reported that the primary water stress corrosion cracking (PWSCC) has been occurred at the nickel based alloy weld components such as steam generator safe end weld, reactor vessel safe end weld, and so on, in PWR. Defect detection and sizing is important in order to ensure the reliable operation and life extension of nuclear power plants. In the reactor vessel safe end weld, it was impossible to measure crack depth of PWSCC. The crack was detected in the axial direction of the safe end weld. Furthermore, the crack had some features such as shallow, large aspect ratio (ratio of crack depth and length), sharp geometry of crack tip, and so on. Therefore, development and improvement of defect detection and sizing capabilities for ultrasonic inspection technique is required. Phased array UT technique was applied to nickel based alloy weld specimen with SCC cracks. From the experimental results, good accuracy of crack depth sizing by phased array UT for the inside inspection was shown. From these results, UT procedure for crack depth sizing was verified. Therefore, effectiveness of phased array UT for crack depth sizing in the nickel based alloy welds was shown. (author)

Stiffness-constant variation in nickel-based alloys: Experiment and theory

International Nuclear Information System (INIS)

Hennion, M.; Hennion, B.

1979-01-01

Recent measurements of the spin-wave stiffness constant in several nickel alloys at various concentrations are interpreted within a random-phase approximation, coherent-potential approximation (RPA-CPA) band model which uses the Hartree-Fock approximation to treat the intraatomic correlations. We give a theoretical description of the possible impurity states in the Hartree-Fock approximation. This allows the determination of the Hartree-Fock solutions which can account for the stiffness-constant behavior and the magnetic moment on the impurity for all the investigated alloys. For alloys such as NiCr, NiV, NiMo, and NiRu, the magnetizations of which deviate from the Slater-Pauling curve, our determination does not correspond to previous works and is consequently discussed. The limits of the model appear mainly due to local-environment effects; in the case of NiMn, it is found that a ternary-alloy model with some Mn atoms in the antiferromagnetic state can account for both stiffness-constant and magnetization behaviors

Acoustic emission analysis coupled with thermogravimetric experiments dedicated to high temperature corrosion studies on metallic alloys

International Nuclear Information System (INIS)

Serris, Eric; Al Haj, Omar; Peres, Veronique; Cournil, Michel; Kittel, Jean; Grosjean, Francois; Ropital, Francois

2014-01-01

High temperature corrosion of metallic alloys (like iron, nickel, zirconium alloys) can damage equipment of many industrial fields (refinery, petrochemical, nuclear..). Acoustic emission (AE) is an interesting method owing to its sensitivity and its non-destructive aspect to quantify the level of damage in use of these alloys under various environmental conditions. High temperature corrosive phenomena create stresses in the materials; the relaxation by cracks of these stresses can be recorded and analyzed using the AE system. The goal of our study is to establish an acoustic signals database which assigns the acoustic signals to the specific corrosion phenomena. For this purpose, thermogravimetric analysis (TGA) is coupled with acoustic emission (AE) devices. The oxidation of a zirconium alloy, zircaloy-4, is first studied using thermogravimetric experiment coupled to acoustic emission analysis at 900 C. An inward zirconium oxide scale, preliminary dense, then porous, grow during the isothermal isobaric step. The kinetic rate increases significantly after a kinetic transition (breakaway). This acceleration occurs with an increase of acoustic emission activity. Most of the acoustic emission bursts are recorded after the kinetic transition. Acoustic emission signals are also observed during the cooling of the sample. AE numerical treatments (using wavelet transform) completed by SEM microscopy characterizations allows us to distinguish the different populations of cracks. Metal dusting represents also a severe form of corrosive degradation of metal alloy. Iron metal dusting corrosion is studied by AE coupled with TGA at 650 C under C 4 H 10 + H 2 + He atmosphere. Acoustic emission signals are detected after a significant increase of the sample mass.