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Sample records for zirconium alloys resistant

  1. Zirconium alloy barrier having improved corrosion resistance

    International Nuclear Information System (INIS)

    Adamson, R.B.; Rosenbaum, H.S.

    1983-01-01

    A nuclear fuel element for use in the core of a nuclear reactor has a composite cladding container having a substrate and a dilute zirconium alloy liner bonded to the inside surface of the substrate. The dilute zirconium alloy liner forms about 1 to about 20 percent of the thickness of the cladding and is comprised of zirconium and a metal selected from the group consisting of iron, chromium, iron plus chromium, and copper. The dilute zirconium alloy liner shields the substrate from impurities or fission products from the nuclear fuel material and protects the substrate from stress corrosion and stress cracking. The dilute zirconium alloy liner displays greater corrosion resistance, especially to oxidation by hot water or steam than unalloyed zirconium. The substrate material is selected from conventional cladding materials, and preferably is a zirconium alloy. (author)

  2. 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

  3. Highly corrosion resistant zirconium based alloy for reactor structural material

    International Nuclear Information System (INIS)

    Ito, Yoichi.

    1996-01-01

    The alloy of the present invention is a zirconium based alloy comprising tin (Sn), chromium (Cr), nickel (Ni) and iron (Fe) in zirconium (Zr). The amount of silicon (Si) as an impurity is not more than 60ppm. It is preferred that Sn is from 0.9 to 1.5wt%, that of Cr is from 0.05 to 0.15wt%, and (Fe + Ni) is from 0.17 to 0.5wt%. If not less than 0.12wt% of Fe is added, resistance against nodular corrosion is improved. The upper limit of Fe is preferably 0.40wt% from a view point of uniform suppression for the corrosion. The nodular corrosion can be suppressed by reducing the amount of Si-rich deposition product in the zirconium based alloy. Accordingly, a highly corrosion resistant zirconium based alloy improved for the corrosion resistance of zircaloy-2 and usable for a fuel cladding tube of a BWR type reactor can be obtained. (I.N.)

  4. 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.)

  5. 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)

  6. PLUTONIUM-ZIRCONIUM ALLOYS

    Science.gov (United States)

    Schonfeld, F.W.; Waber, J.T.

    1960-08-30

    A series of nuclear reactor fuel alloys consisting of from about 5 to about 50 at.% zirconium (or higher zirconium alloys such as Zircaloy), balance plutonium, and having the structural composition of a plutonium are described. Zirconium is a satisfactory diluent because it alloys readily with plutonium and has desirable nuclear properties. Additional advantages are corrosion resistance, excellent fabrication propenties, an isotropie structure, and initial softness.

  7. 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

  8. Effects of alloying elements on nodular and uniform corrosion resistance of zirconium-based alloys

    International Nuclear Information System (INIS)

    Abe, Katsuhiro

    1992-01-01

    The effects of alloying and impurity elements (tin, iron, chromium, nickel, niobium, tantalum, oxygen, aluminum, carbon, nitrogen, silicon, and phosphorus) on the nodular and uniform corrosion resistance of zirconium-based alloys were studied. The improving effect of iron, nickel and niobium in nodular corrosion resistance were observed. The uniform corrosion resistance was also improved by nickel, niobium and tantalum. The effects of impurity elements, nitrogen, aluminum and phosphorus were negligibly small but increasing the silicon content seemed to improve slightly the uniform corrosion resistance. Hydrogen pick-up fraction were not changed by alloying and impurity elements except nickel. Nickel addition increased remarkably hydrogen pick-up fraction. Although the composition of secondary precipitates changed with contents of alloying elements, the correlation of composition of secondary precipitates to corrosion resistance was not observed. (author)

  9. Components made of corrosion-resistent zirconium alloy and method for its production

    International Nuclear Information System (INIS)

    Hanneman, R.E.; Urquhart, A.W.; Vermilyea, D.A.

    1977-01-01

    The invention deals with a method to increase the resistance of zirconium alloys to blister corrosion which mainly occurs in boiling-water nuclear reactors. According to the method described, the surface of the alloy body is coated with a thin film of a suitable electronically conducting material. Gold, silver, platinum, nickel, chromium, iron and niobium are suitable as coating materials. The invention is more closely explained by means of examples. (GSC) [de

  10. 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.)

  11. 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

  12. 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.

  13. 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

  14. Electroless deposition process for zirconium and zirconium alloys

    Science.gov (United States)

    Donaghy, Robert E.; Sherman, Anna H.

    1981-01-01

    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.

  15. 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

  16. 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

  17. 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.

  18. Thermofluency in zirconium alloys

    International Nuclear Information System (INIS)

    Orozco M, E.A.

    1976-01-01

    A summary is presented about the theoretical and experimental results obtained at present in thermofluency under radiation in zirconium alloys. The phenomenon of thermofluency is presented in a general form, underlining the thermofluency at high temperature because this phenomenon is similar to the thermofluency under radiation, which ocurrs in zirconium alloys into the operating reactor. (author)

  19. 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

  20. 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)

  1. Influence of impurities and ion surface alloying on the corrosion resistance of E110 alloy

    International Nuclear Information System (INIS)

    Kalin, B. A.; Volkov, N. V.; Valikov, R. A.; Novikov, V. V.; Markelov, V. A.; Pimenov, Yu. V.

    2013-01-01

    The corrosion resistance of zirconium alloys depends on their structural-phase state, the type of core coolant and operating factors. The formation of a protective oxide film on the zirconium alloys is sensitive to the content of impurity atoms present in the charge base of alloys and accumulating in them in the manufacture of products. The impurity composition of the initial zirconium is determined by the method of its manufacture and generally remains unchanged in the products, deter-mining their properties, including their corrosion resistance. An increased content of impurities (C, N, Al, Mo, Fe) both individually and in their combination negatively affects the corrosion resistance of zirconium and its alloys. One of the potentially effective methods to increase the protective properties of oxide films on zirconium alloys is a surface alloying using the regime of mixing the atoms of a film, preliminarily coated on the surface, and the atoms of a target. This method makes it possible to form a given structural-phase state in the thin surface layer with unique physicochemical properties and thus to in-crease the corrosion resistance and wear resistance of fuel claddings. In this context, the object of investigation was samples of cladding tubes from alloy E110 with various content of impurity elements (nitrogen, aluminum, and carbon) with the aim to reduce the negative influence of impurities on the corrosion resistance by changing the structural-phase state of the surface layer of fuel claddings and fuel assembly components with alloying in the regime of ion mixing of atoms

  2. The development of zirconium alloy and its manufacturing

    International Nuclear Information System (INIS)

    Yuan Gaihuan; Yue Qiang

    2015-01-01

    Nuclear power which acts as one of low-carbon energy resources is the most realistic in large-scale application. It is also the preferred choice for many countries to develop energy resources and optimize its structure. Zirconium alloy is a key structural material for nuclear power plant fuel assemblies and cladding tubes of zirconium alloy are often referred as the first safeguard to nuclear power safety. With the development of nuclear power, three kinds of zirconium alloys Zr-Sn, Zr-Nb, Zr-Sn-Nb and with the representative products of Zr-4, M5, Zirlo respectively are developed and widely applied. Because of its severe operating environment and influence to nuclear safety, the requirements to zirconium alloys for physical and chemical properties, nuclear capability, tolerance and surface quality are very strict. The in-depth research and its manufacture capability become one of the main barriers for many countries who are developing the nuclear energy. In recent years, a stated-owned company, State Nuclear Bao Ti Zirconium Industry Company ('SNZ' for short) as well as National R and D Center for Nuclear Grade Zirconium material, is founded to meet the requirement of the rapid development of China's nuclear power industry. SNZ is dedicated for the fabrication and the research of nuclear grade zirconium products. After the successful completion of technology transfer of manufacturing for production chain and fully grasped of the manufacturing technology for the nuclear grade zirconium sponge through zirconium alloy tube, rod and strip products. National R and D Center for Nuclear Grade Zirconium material is cooperating with universities, nuclear energy research and design institutes and the owners of nuclear power plant to develop new zirconium alloy of self-owned brand. Through the selection of components, in-process testing and product inspection, four kinds of new zirconium alloys owns better performance than currently commercialized M5, Zirlo etc

  3. Applications for zirconium and columbium alloys

    International Nuclear Information System (INIS)

    Condliff, A.F.

    1986-01-01

    Currently, zirconium and columbium are used in a wide range of applications, overlapping only in the field of corrosion control. As a construction material, zirconium is primarily used by the nuclear power industry. The use of zirconium in the chemical processing industry (CPI) is, however, increasing steadily. Columbian alloys are primarily applied as superconducting alloys for research particle accelerators and fusion generators as well as in magnetic resonance imaging for medical diagnosis

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

    Science.gov (United States)

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

    2017-02-01

    To analyze and evaluate imaging artefacts induced by zirconium, titanium and titanium-zirconium alloy dental implants. 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. 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. 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.

  5. 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

  6. 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

  7. 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.

  8. 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

  9. 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)

  10. 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.

  11. 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.)

  12. 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

  13. 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

  14. 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.

  15. 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

  16. 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.)

  17. 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

  18. 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

  19. Corrosion resistance of high-performance materials titanium, tantalum, zirconium

    CERN Document Server

    2012-01-01

    Corrosion resistance is the property of a material to resist corrosion attack in a particular aggressive environment. Although titanium, tantalum and zirconium are not noble metals, they are the best choice whenever high corrosion resistance is required. The exceptionally good corrosion resistance of these high–performance metals and their alloys results from the formation of a very stable, dense, highly adherent, and self–healing protective oxide film on the metal surface. This naturally occurring oxide layer prevents chemical attack of the underlying metal surface. This behavior also means, however, that high corrosion resistance can be expected only under neutral or oxidizing conditions. Under reducing conditions, a lower resistance must be reckoned with. Only very few inorganic and organic substances are able to attack titanium, tantalum or zirconium at ambient temperature. As the extraordinary corrosion resistance is coupled with an excellent formability and weldability these materials are very valua...

  20. 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.

  1. 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.

  2. 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

  3. 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

  4. 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)

  5. 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.

  6. 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)

  7. 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.

  8. 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)

  9. 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

  10. 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.

  11. 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

  12. 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)

  13. 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.

  14. 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.)

  15. 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.

  16. 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)

  17. 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.

  18. 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.

  19. 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

  20. 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)

  1. 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.

  2. Advanced Corrosion-Resistant Zr Alloys for High Burnup and Generation IV Applications

    International Nuclear Information System (INIS)

    Arthur Motta; Yong Hwan Jeong; R.J. Comstock; G.S. Was; Y.S. Kim

    2006-01-01

    The objective of this collaboration between four institutions in the US and Korea is to demonstrate a technical basis for the improvement of the corrosion resistance of zirconium-based alloys in more extreme operating environments (such as those present in severe fuel duty, cycles high burnup, boiling, aggressive chemistry) and to investigate the feasibility (from the point of view of corrosion rate) of using advanced zirconium-based alloys in a supercritical water environment

  3. Improvements in zirconium alloy corrosion resistance

    International Nuclear Information System (INIS)

    Kilp, G.R.; Thornburg, D.R.; Comstock, R.J.

    1990-01-01

    The corrosion rates of a series of Zircaloy 4 and Zr-Nb alloys were evaluated in long-term (exceeding 500 days in some cases) autoclave tests. The testing was done at various conditions including 633 K (680 F) water, 633 K (650 F) water, 633 k (680 F) lithiated water (70 PPM/0.01 molal lithium), and 673 K (750 F) steam. Materials evaluated are from the following three groups: (1) standard Zircaloy 4; (2) Zircaloy 4 with tightened controls on chemistry limits and heat-treatment history; and (3) Zr-Nb alloys. To optimize the corrosion resistance of the Zircaloy 4 material, the effects of specific chemistry controls (tighter limits on nitrogen, oxygen, silicon, carbon and tin) were evaluated. Also the effects of the thermal history, as measured by integrated annealing of ''A'' time were determined. The ''A'' times ranged from 0.1x10 -18 (h) to 46x10 -18 (h). A material referred to as ''Improved Zircaloy 4'', having optimized chemistry and ''A'' time levels for reduced corrosion, has been developed and tested. This material has a reduced and more uniform corrosion rate compared to the prior Zircaloy 4 material. Alternative alloys were also evaluated for potential improvement in cladding corrosion resistance. ZIRLO TM material was chosen for development and has been included in the long-term corrosion testing. Demonstration fuel assemblies using ZIRLO cladding are now operating in a commercial reactor. The results for the various test conditions and compositions are reported and the relative corrosion characteristics summarized. Based on the BR-3 data, there is a ranking correspondence between in-reactor corrosion and autoclave testing in lithiated water. In particular, the ZIRLO material has significantly improved relative corrosion resistance in the lithiated water tests. Reduced Zircaloy-4 corrosion rates are also obtained from the tighter controls on the chemistry (specifically lower tin, nitrogen, and carbon; higher silicon; and reduced oxygen variability) and ''A

  4. 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.

  5. Corrosion resistance of zirconium: general mechanisms, behaviour in nitric acid

    International Nuclear Information System (INIS)

    Pinard Legry, G.

    1990-01-01

    Corrosion resistance of zirconium results from the strong affinity of this metal for oxygen; as a result a thin protective oxide film is spontaneously formed in air or aqueous media, its thickness and properties depending on the physicochemical conditions at the interface. This film passivates the underlying metal but obviously if the passive film is partially or completely removed, localised or generalised corrosion phenomena will occur. In nitric acid, this depassivation may be chemical (fluorides) or mechanical (straining, creep, fretting). In these cases it is useful to determine the physicochemical conditions (concentration, temperature, potential, stress) which will have to be observed to use safely zirconium and its alloys in nitric acid solutions [fr

  6. 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)

  7. 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

  8. Erosion resistance of composite materials on titanium, zirconium and aluminium nitride base under the electron beam effect

    International Nuclear Information System (INIS)

    Verkhoturov, A.D.; Kuzenkova, M.A.; Slutskin, M.G.; Kravchuk, L.A.

    1977-01-01

    Erosion resistance of composites based on nitrides of titanium, zirconium and aluminium to spark and electron beam processing has been studied. The erosion resistance in spark processing is shown to depend on specific electric resistance of the alloys. TiN-AlN and ZrN-AlN alloys containing more than 70% AlN (with specific electric resistance more than 10 6 -10 7 ohm/cm) caot be processed by spark method. It is shown that erosion of the composites by an electron beam depends primarily on the rate of evaporation of the components

  9. In situ Raman Spectroscopy of Oxide Films on Zirconium Alloy in Simulated PWR Primary Water Condition

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Tae Ho; Choi, Kyoung Joon; Yoo, Seung Chang; Kim, Ji Hyun [UNIST, Ulsan (Korea, Republic of)

    2016-05-15

    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 pretransition zirconium oxide in high-temperature water chemistry. In this work, Raman spectroscopy was used for in situ investigations for characterizing the phase of zirconium oxide. In situ Raman spectroscopy is a well-suited technique for investigating in detail the characteristics of oxide films in a high-temperature corrosion environment. In previous studies, an in situ Raman system was developed for investigating the oxides on nickel-based alloys and low alloy steels in high-temperature water environment. Also, the early stage oxidation behavior of zirconium alloy with different dissolved hydrogen concentration environments in high temperature water was treated in the authors' previous study. In this study, a specific zirconium alloy was oxidized and investigated with in situ Raman spectroscopy for 100 d oxidation, which is close to the first transition time of the zirconium alloy oxidation. The ex situ investigation methods such as transmission electron microscopy (TEM) and energy dispersive X-ray spectroscopy (EDS) were used to further characterize the zirconium oxide structure. As oxidation time increased, the Raman peaks of tetragonal zirconium oxide were merged or became weaker. However, the monoclinic zirconium oxide peaks became distinct. The tetragonal zirconium oxide was just found near the O/M interface and this could explain the Raman spectra difference between the 30 d result and others.

  10. 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

  11. 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)

  12. 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

  13. 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)

  14. 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.)

  15. 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

  16. Studying titanium-molybdenum-zirconium alloys of increased corrosion resistance in acid solutions

    International Nuclear Information System (INIS)

    Tomashov, N.D.; Kazarin, V.I.; Mikheev, V.S.; Goncharenko, B.A.; Sigalovskaya, T.M.; Kalyanova, M.P.

    1977-01-01

    New promising Ti-Mo-Nb-Zr system alloys, possessing good workability and a high corrosion resistance in non-oxidizing solutions of acids, have been developed. The alloys may be recommended as structural materials for equipment operating in severely agressive acid media, such as hydrochloric, sulphuric and phosphoric acids. The corrosion resistance of alloys of the above system in solutions of H 2 SO 4 , HCl and H 3 PO 4 acids may be maximized by increasing the overall alloying to 42% (keeping the ratio of the alloying components Mo/Nb/Zr=4/1/1 unchanged), while retaining sufficiently good plasticity and workability

  17. 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

  18. 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

  19. 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)

  20. 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.

  1. 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

  2. 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

  3. 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

  4. 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)

  5. 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)

  6. 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

  7. 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.

  8. 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

  9. 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.)

  10. 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.)

  11. High corrosion-resistant fuel spacers

    International Nuclear Information System (INIS)

    Yoshida, Toshimi; Takase, Iwao; Ikeda, Shinzo; Masaoka, Isao; Nakajima, Junjiro.

    1986-01-01

    Purpose: To enable manufacturing BWR fuel spacers by prior-art production process, using a zirconium-base alloy having very excellent corrosion resistance. Method: A highly improved nodular-resistant, corrosion-resistant zirconium alloy is devised by adding a slight amount of niobium, titanium and vanadium to zircaloy, of which fuel spacers are produced. That is, there can be obtained an alloy having much more excellent nodular resistance than conventional zircaloy, and free from a large change in plasticity, workability, and weldability, by adding to zirconium about 1.5 % of tin, about 0.15 % of iron, about 0.05 % of chromium, about 0.05 % of nickel, and 0.05 to 0.5 % of at least one or two kinds of niobium, titanium and vanadium. Using this zirconium-base alloy can manufacture fuel spacers by the same manufacturing process, thus improving economy and reliability. (Kamimura, M.)

  12. 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

  13. 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

  14. 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

  15. Accelerated irradiation growth of zirconium alloys

    International Nuclear Information System (INIS)

    Griffiths, M.; Gilbert, R.W.; Fidleris, V.

    1989-01-01

    This paper discusses how sponge zirconium and Zr-2.5 wt% Nb, Zircaloy, or Excel alloys all exhibit accelerated irradiation growth compared with high-purity crystal-bar zirconium for irradiation temperatures between 550 to 710 K and fluences between 0.1 to 10 x 10 25 n · m -2 (E > 1 MeV). There is generally an incubation period or fluence before the onset of accelerated or breakaway growth, which is dependent on the particular material being irradiated, its metallurgical condition before irradiation, and the irradiation temperature. Transmission electron microscopy has shown that there is a correlation between accelerated irradiation growth and the appearance of c-component vacancy loops on basal planes. Measurements in some specimens indicate that a significant fraction of the strain can be directly attributed to the loops themselves. There is considerable evidence to show that their formation is dependent both on the specimen purity and on the irradiation temperature. Materials that have a high interstitial-solute content contain c-component loops and exhibit high growth rates even at low fluences ( 2 :5 n · m -2 , E > 1 MeV). For sponge zirconium and the Zircaloys, c-component loop formation and the associated acceleration of growth (breakaway) during irradiation occurs because the intrinsic interstitial solute (mainly, oxygen, carbon and nitrogen) in the zirconium matrix is supplemented by interstitial iron, chromium, and nickel from the radiation-induced dissolution of precipitates. (author)

  16. Hardening of niobium alloys at precrystallization annealing

    International Nuclear Information System (INIS)

    Vasil'eva, E.V.; Pustovalov, V.A.

    1989-01-01

    Niobium base alloys were investigated. It is shown that precrystallization annealing of niobium-molybdenum, niobium-vanadium and niobium-zirconium alloys elevates much more sufficiently their resistance to microplastic strains, than to macroplastic strains. Hardening effect differs sufficiently for different alloys. The maximal hardening is observed for niobium-vanadium alloys, the minimal one - for niobium-zirconium alloys

  17. 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.

  18. 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)

  19. 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.

  20. 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

  1. Solute redistribution studies in oxidised zirconium alloys

    Energy Technology Data Exchange (ETDEWEB)

    Khera, S K; Kale, G B; Gadiyar, H S [Bhabha Atomic Research Centre, Bombay (India). Metallurgy Div.

    1977-01-01

    Electron microprobe studies on solute distribution in oxide layers and in the regions near oxide metal interface have been carried out in the case of zircaloy-2 and zirconium binary alloys containing niobium, tin, iron, copper, chromium and nickel and oxidised in steam at 550 deg C. In the case of alloys having higher oxidation rates, the oxide of solute element was found to dissolve in ZrO/sub 2/ without any composition variation. However, for solute addition with limited solubility like Cr, Cu and Fe, solute enrichment at metal/oxide interface and depletion of the same matrix has been observed. The intensity profiles for nickel distribution were also found to be identical to Fe or Cr distribution. The mode of solute distribution has been discussed in relation to oxidation behaviour of these alloys.

  2. 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

  3. 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.

  4. 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

  5. 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.)

  6. Analysis of zirconium and nickel based alloys and zirconium oxides by relative and internal monostandard neutron activation analysis methods

    International Nuclear Information System (INIS)

    Shinde, Amol D.; Acharya, Raghunath; Reddy, Annareddy V. R.

    2017-01-01

    The chemical characterization of metallic alloys and oxides is conventionally carried out by wet chemical analytical methods and/or instrumental methods. Instrumental neutron activation analysis (INAA) is capable of analyzing samples nondestructively. As a part of a chemical quality control exercise, Zircaloys 2 and 4, nimonic alloy, and zirconium oxide samples were analyzed by two INAA methods. The samples of alloys and oxides were also analyzed by inductively coupled plasma optical emission spectroscopy (ICP-OES) and direct current Arc OES methods, respectively, for quality assurance purposes. The samples are important in various fields including nuclear technology. Samples were neutron irradiated using nuclear reactors, and the radioactive assay was carried out using high-resolution gamma-ray spectrometry. Major to trace mass fractions were determined using both relative and internal monostandard (IM) NAA methods as well as OES methods. In the case of alloys, compositional analyses as well as concentrations of some trace elements were determined, whereas in the case of zirconium oxides, six trace elements were determined. For method validation, British Chemical Standard (BCS)-certified reference material 310/1 (a nimonic alloy) was analyzed using both relative INAA and IM-NAA methods. The results showed that IM-NAA and relative INAA methods can be used for nondestructive chemical quality control of alloys and oxide samples

  7. Analysis of zirconium and nickel based alloys and zirconium oxides by relative and internal monostandard neutron activation analysis methods

    Energy Technology Data Exchange (ETDEWEB)

    Shinde, Amol D.; Acharya, Raghunath; Reddy, Annareddy V. R. [Bhabha Atomic Research Centre, Mumbai (India)

    2017-04-15

    The chemical characterization of metallic alloys and oxides is conventionally carried out by wet chemical analytical methods and/or instrumental methods. Instrumental neutron activation analysis (INAA) is capable of analyzing samples nondestructively. As a part of a chemical quality control exercise, Zircaloys 2 and 4, nimonic alloy, and zirconium oxide samples were analyzed by two INAA methods. The samples of alloys and oxides were also analyzed by inductively coupled plasma optical emission spectroscopy (ICP-OES) and direct current Arc OES methods, respectively, for quality assurance purposes. The samples are important in various fields including nuclear technology. Samples were neutron irradiated using nuclear reactors, and the radioactive assay was carried out using high-resolution gamma-ray spectrometry. Major to trace mass fractions were determined using both relative and internal monostandard (IM) NAA methods as well as OES methods. In the case of alloys, compositional analyses as well as concentrations of some trace elements were determined, whereas in the case of zirconium oxides, six trace elements were determined. For method validation, British Chemical Standard (BCS)-certified reference material 310/1 (a nimonic alloy) was analyzed using both relative INAA and IM-NAA methods. The results showed that IM-NAA and relative INAA methods can be used for nondestructive chemical quality control of alloys and oxide samples.

  8. 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...

  9. 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.

  10. Enhanced antimicrobial properties, cytocompatibility, and corrosion resistance of plasma-modified biodegradable magnesium alloys.

    Science.gov (United States)

    Zhao, Ying; Jamesh, Mohammed Ibrahim; Li, Wing Kan; Wu, Guosong; Wang, Chenxi; Zheng, Yufeng; Yeung, Kelvin W K; Chu, Paul K

    2014-01-01

    Magnesium alloys are potential biodegradable materials and have received increasing attention due to their outstanding biological performance and mechanical properties. However, rapid degradation in the physiological environment and potential toxicity limit clinical applications. Recently, special magnesium-calcium (Mg-Ca) and magnesium-strontium (Mg-Sr) alloys with biocompatible chemical compositions have been reported, but the rapid degradation still does not meet clinical requirements. In order to improve the corrosion resistance, a rough, hydrophobic and ZrO(2)-containing surface film is fabricated on Mg-Ca and Mg-Sr alloys by dual zirconium and oxygen ion implantation. Weight loss measurements and electrochemical corrosion tests show that the corrosion rate of the Mg-Ca and Mg-Sr alloys is reduced appreciably after surface treatment. A systematic investigation of the in vitro cellular response and antibacterial capability of the modified binary magnesium alloys is performed. The amounts of adherent bacteria on the Zr-O-implanted and Zr-implanted samples diminish remarkably compared to the unimplanted control. In addition, significantly enhanced cell adhesion and proliferation are observed from the Zr-O-implanted sample. The results suggest that dual zirconium and oxygen ion implantation, which effectively enhances the corrosion resistance, in vitro biocompatibility and antimicrobial properties of Mg-Ca and Mg-Sr alloys, provides a simple and practical means to expedite clinical acceptance of biodegradable magnesium alloys. Copyright © 2013 Acta Materialia Inc. Published by Elsevier Ltd. All rights reserved.

  11. 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

  12. 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

  13. 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

  14. 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.

  15. 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

  16. Corrosion resistance of the niobium-zirconium-oxygen alloys in the molten lithium

    International Nuclear Information System (INIS)

    Arakelov, A.G.; Vavilova, V.V.; Gekov, A.F.; Zel'tser, A.M.

    1977-01-01

    Phase behaviour of Nb-Zr-O system alloys after thermal treatment at 1500 deg and 500 deg C has been studied in the concentration range up to 6 at.% Zr and 6 at.% O. Alloys annealed at 1500 deg C, so that the ratio Zr:O was 1:2, displayed intercrystalline corrosion in lithium environment, whereas after annealing at 500 deg C the corrosion was largely transcrystalline. Lithium penetration into these alloys which is much slower than that into Nb-O alloys, results, as in the binary system, in lower microhardness and higher specific electrical resistance

  17. 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

  18. Study on the improvement of the properties of Zr alloys

    International Nuclear Information System (INIS)

    Kim, Young Suk; Han, Jung Ho; Jeong, Yong Hwan; Lee, Duk Hyun; Park, Gi Sung; Hong, Jun Hwa; Park, Ji Yun; No, Gae Ho

    1992-01-01

    1) The objective of this study is to develop the corrosion resistant zirconium base alloys. In order to achieve this goal, this year's activities have focused on the guidelines for the corrosion resistant zirconium alloy design, the manufacturing of the sheets of zirconium base alloys and finally the characterization of the NAZAs (New Alternate Zirconium alloys). The main results from this study can be summarized as follows: 2) Based on the evaluation of the role of alloying elements, i.e., Nb, Sn, Fe, Cr, and etc, as many as 23 different kinds of the NAZAs were preliminarily designed. 3) The 3 kinds of the NAZAs-Lot 15, 22 and 23 manufactured into a sheet though a series of manufacturing procedures. 4) The microstructures, hardness and the corrosion performances of 3 kinds of NAZAs were investigated. (Author)

  19. 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.)

  20. 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.

  1. 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

  2. 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

  3. 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

  4. Investigation on the corrosion resistance of zirconium in nitric acid

    International Nuclear Information System (INIS)

    Fauvet, P.; Mur, P.

    1990-01-01

    Zirconium in nitric solutions exhibits an excellent corrosion resistance in the passive state, and a mediocre corrosion resistance in the unpassive state with risk of stress corrosion cracking. Results of the influence of some parameters (medium, potential, temperature, stress, friction, metallurgical structure and surface state) on zirconium passivation are presented. Zirconium remains passive in a large range of HNO 3 concentration (at least up to 14.4N), in the presence of oxidizing ions (Cr 4 , Ce 4 ), in a spent fuel dissolution solution. Zirconium is depassived by friction at high speed and pressure, by platinum coupling in boiling 14.4N HNO 3 with or without stress, or by imposed deformation speed under high potential. (A.B.)

  5. 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

  6. 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

  7. 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.)

  8. Oxidation kinetics and auger microprobe analysis of some oxidized zirconium alloys

    International Nuclear Information System (INIS)

    Ploc, R.A.

    1989-01-01

    Oxidation kinetics at 300 o C in dry oxygen of 0.5 wt% binary alloys of iron, nickel, and chromium in zirconium were determined for several surface preparations. Further, chemical profiles of the oxides as they existed on the matrix and on the precipitates were obtained by sputtering and Auger electron analysis. The appearance of 'breakaway' oxidation was controlled by the surface finish of the alloy, a variable that could be used to eliminate the phenomenon for all alloys except the Zr/Ni binary, which required β-quenching to accomplish the same purpose. (author)

  9. 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.

  10. 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.

  11. Effect of a ZrO{sub 2} coating deposited by the sol–gel method on the resistance of FeCrAl alloy in high-temperature oxidation conditions

    Energy Technology Data Exchange (ETDEWEB)

    Chęcmanowski, Jacek Grzegorz, E-mail: jacek.checmanowski@pwr.wroc.pl [Wrocław University of Technology, Faculty of Chemistry, Wybrzeże Wyspiańskiego 27, 50-370 Wrocław (Poland); Szczygieł, Bogdan, E-mail: bogdan.szczygiel@pwr.wroc.pl [Wrocław University of Technology, Faculty of Chemistry, Wybrzeże Wyspiańskiego 27, 50-370 Wrocław (Poland)

    2013-05-15

    One-, three- and five-layer protective ZrO{sub 2} coatings were deposited on a FeCrAl alloy base by the sol–gel method. A zirconium(IV) isopropoxide isopropanol complex was used as the zirconium precursor. It has been shown that zirconium in the amount of 0.3–0.5 wt.% improves the resistance of FeCrAl alloy in high-temperature oxidation conditions (in air at T = 1060 °C for t = 2400 h). Even a very low Zr content affects the morphology, porosity and composition of the forming scale (SEM, EDS). An analysis of the chemical composition of the material after oxidation indicated to-core Zr diffusion. The presence of zirconium prevents catastrophic corrosion of the FeCrAl alloy during oxidation. In the case of the alloy without the reactive element (Zr) this type of corrosion occurred after about 1800 h. The oxidation of the FeCrAl alloy covered with ZrO{sub 2} coatings proceeds in three stages. In the first stage, lasting about 50 h, the mass of the sample grows rapidly, then for 700 h the mass changes minimally and in the third stage the oxidation proceeds according to a parabolic dependence. The presence of Zr on the surface of the FeCrAl alloy significantly contributes to the protective effect of the coatings. - Highlights: ► Multilayer ZrO{sub 2} coatings were deposited on FeCrAl alloy by sol–gel method. ► Study of alloy composition indicates to-core Zr diffusion in high temperature. ► Even very low content affects morphology and porosity of forming scale. ► Zirconium improves the resistance of FeCrAl alloy in high temperature conditions. ► Presence of ZrO{sub 2} prevents catastrophic corrosion of FeCrAl alloy during oxidation.

  12. 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)

  13. 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

  14. 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

  15. 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

  16. 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)

  17. The corrosion resistance of Zr-Nb and Zr-Nb-Sn alloys in high-temperature water and steam

    International Nuclear Information System (INIS)

    Dalgaard, S.B.

    1960-03-01

    An alloy of reactor-grade sponge zirconium-2.5 wt. % niobium was exposed to water and steam at high temperature. The corrosion was twice that of Zircaloy-2 while hydrogen pickup was found to be equal to that of Zircaloy-2. Ternary additions of tin to this alloy in the range 0.5-1.5 had no effect on the corrosion resistance in water at 315 o C up to 100 days. At higher temperatures, tin increased the corrosion, the effect varying with temperature. Heat treatment of the alloys was shown to affect corrosion resistance. (author)

  18. The corrosion resistance of Zr-Nb and Zr-Nb-Sn alloys in high-temperature water and steam

    Energy Technology Data Exchange (ETDEWEB)

    Dalgaard, S B

    1960-03-15

    An alloy of reactor-grade sponge zirconium-2.5 wt. % niobium was exposed to water and steam at high temperature. The corrosion was twice that of Zircaloy-2 while hydrogen pickup was found to be equal to that of Zircaloy-2. Ternary additions of tin to this alloy in the range 0.5-1.5 had no effect on the corrosion resistance in water at 315{sup o}C up to 100 days. At higher temperatures, tin increased the corrosion, the effect varying with temperature. Heat treatment of the alloys was shown to affect corrosion resistance. (author)

  19. 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)

  20. 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

  1. The Method of Measured Electrical Resistivity in Studying Phase Transformations in Zr1Nb Alloy

    International Nuclear Information System (INIS)

    Gritsina, V.M.; Klimenko, S.P.; Chernyaeva, T.P.

    2006-01-01

    The paper systematically arranges and analyzes the data on the methods of research into α ↔ β transformation process in zirconium alloys, as well as capabilities and information provided by each method. A special emphasis is put on the method of measured electrical resistivity. The authors also present the results of their own research into α ↔ β transformation process in Zr1Nb alloy (in the material of Zr+1% Nb tubing produced in Ukraine from calciothermal zirconium). The ρ →T curve was used to define the maximum and minimum values for transformation temperatures. Combined processing of the phase data on Zr+1% Nb found in literature and obtained from measured resistivity suggests that transformation process happens in several stages. The maximum value on the ρ → T curve corresponds to the beginning of stage 3, whereas the minimum - to its completion; as suggested by the pooled data, accounts for over 95% of the total volume of the material

  2. Electrochemical heterogeneity and corrosion resistance of a welded titanium-zirconium joint

    International Nuclear Information System (INIS)

    Polyakov, S.G.; Goncharov, A.B.; Onoprienko, L.M.; Smiyan, O.D.

    1992-01-01

    The electrochemical behavior and corrosion resistance of various welded joints of zirconium alloy N-2.5 with commercial titanium VT1 made by the argon-arc method are studied. Electrochemical heterogeneity is studied by measuring the distribution of potentials over the surface, galvanic currents, and recording of polarization curves for different zones of a welded joint in 5% sulfuric acid solution at 340 K. It is established that electrochemical heterogeneity of the zones of an N-2.5 + VT1 welded joint leads to acceleration of the cathodic process in a welded joint and the anodic process along the fusion line from the titanium direction where the greatest hydrogenation of the metal and corrosion damage is correspondingly observed

  3. 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)

  4. 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.

  5. 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.

  6. Determination of hydrogen in zirconium and its alloys by melt extraction under carrier gas flow using thermal conductivity cell as detector

    International Nuclear Information System (INIS)

    Akhtar, J.; Ahmed, M.; Mohammad, B.; Jan, S.; Waqar, F.

    1987-06-01

    In the production of zirconium metal and its alloys the presence of hydrogen impurity affects mechanical and corrosion resistance properties of the product. Therefore, determination of hydrogen contents of the product is necessary. Conditions for its analysis by melt extraction under carrier gas stream using thermal conductivity cell as detector were studied and optimised. The method is capable of measuring hydrogen impurity in parts per million range. (author)

  7. 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

  8. 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

  9. 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

  10. 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.

  11. 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)

  12. 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)

  13. 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

  14. 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.

  15. 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.

  16. 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

  17. 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)

  18. 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)

  19. 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

  20. Electrical Resistance Alloys and Low-Expansion Alloys

    DEFF Research Database (Denmark)

    Kjer, Torben

    1996-01-01

    The article gives an overview of electrical resistance alloys and alloys with low thermal expansion. The electrical resistance alloys comprise resistance alloys, heating alloys and thermostat alloys. The low expansion alloys comprise alloys with very low expansion coefficients, alloys with very low...... thermoelastic coefficients and age hardenable low expansion alloys....

  1. 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)

  2. 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

  3. 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

  4. 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.

  5. 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.

  6. A computation model for the corrosion resistance of nanocrystalline zirconium metal

    International Nuclear Information System (INIS)

    Zhang Xiyan; Shi Minghua; Liu Nianfu; Wei Yiming; Li Cong; Qiu Shaoyu; Zhang Qiang; Zhang Pengcheng

    2007-01-01

    In this paper a computation model of corrosion rate-grain size of nanocrystalline and ultra-fine zirconium has been presented. The model is based on the Wagner's theory and the electron theory of solids. The conductivity, electronic mean free path and grain size of metal were considered. By this model, the corrosion rate of zirconium metal under different temperature was computed. The results show that the corrosion weight gain and rate constant of nanocrystalline zirconium is lower than that of zirconium with coarse grain size. And the corrosion rate constant and weight gain of nanocrystalline zirconium metal decrease with the decrease of grain size. So the refinement of grain size can remarkably improve the corrosion resistance of zirconium metal. (authors)

  7. 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

  8. 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

  9. 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

  10. 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

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

    Science.gov (United States)

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

    2012-01-01

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

  12. 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

  13. Effects of solutes on damage production and recovery in zirconium

    International Nuclear Information System (INIS)

    Zee, R.H.; Birtcher, R.C.; MacEwen, S.R.; Abromeit, C.

    1986-04-01

    Dilute zirconium-based alloys and pure zirconium were irradiated at 10 K with spallation neutrons at IPNS. Four types of alloys - Zr-Ti, Zr-Sn, Zr-Dy and Zr-Au - each with three concentration levels, were used. Low-temperature resistivity damage rates are enhanced by the presence of any of the four solutes. The greatest enhancement was produced by Au while the least by Dy. Within each alloy group, damage production also increased but at a decreasing rate, with increasing concentration. Post-irradiation annealing experiments, up to 400 K, showed that all four solutes suppress recovery due to interstitial migration, indicative of interstitial trapping by the solutes. Vacancy recovery is also suppressed by the presence of Sn, Dy or Au. The effect of Ti is to shift this stage to lower temperature. No clear correlation between the results with solute size was detected

  14. 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)

  15. Specific heat and electric conductivity of zirconium alloy with 2,5 mass% niobium in the range of phase transitions

    International Nuclear Information System (INIS)

    Roshchupkin, V.V.; Pokrasin, M.A.; Chernov, A.I.; Semashko, N.A.

    1996-01-01

    Experimental investigation of specific heat and electric resistance of zirconium alloy with 2.5 mass% niobium in the range of phase transitions was conducted, using adiabatic calorimeter of original design, characterized by high sensitivity, efficiency and high accuracy. It was revealed that temperature dependence of specific heat was characterized by anomalous growth at 590 deg C, related with (α+β Nb )→(α+β Zr )-transition, and at 810 deg -related with (α+β Zr )→β Zr - transition. Temperature dependence of electric resistance was specific in the region of α+β Zr →β Zr phase transition. It was established that revealed anomalies were connected with high oxygen absorption at high temperatures. 11 refs., 1 fig., 1 tab

  16. 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

  17. Spectrophotometric determination of zirconium in nickel-base alloys with Arsenazo III after separation by froth flotation

    International Nuclear Information System (INIS)

    Sekine, K.; Onishi, H.

    1977-01-01

    0.02-0.1% of zirconium can be determined in nickel alloys by spectrophotometry with Arsenazo III after its separation from the sample solution by means of froth flotation using Arsenazo III and Zephiramine. Nickel, chromium and iron do not interfere. Analysis of standard alloys yielded a standard deviation of 2.2%. (orig.) [de

  18. 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

  19. 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.

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

    Science.gov (United States)

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

    2014-09-01

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

  1. 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

  2. The Development of Corrosion Resistant Zirconium Alloy

    International Nuclear Information System (INIS)

    Abdul-Latief; Noor-Yudhi; Isfandi; Djoko-Kisworo; Pranjono

    2000-01-01

    Corrosion test of Zr alloy consisting of quenching and tempering Zry-2,Zry-4 cast, Zr-1% Nb cast, has been. conducted. In corrosion test, thechanges during β-quenching, tempering and corrosion test at varioustemperature and time in autoclave water medium, can be seen. The treatmentconsisted of heating at 1050 o C for 30 minutes, quenching in water andtempering at 200 o C, 300 o C, 400 o C, 500 o C, 600 o C as well as corrosiontests at 225 o C, 275 o C, 325 o C at 4, 8, 12 hours. Sample preparation forcorrosion test was based on ASTM G-2 procedure, which consisted of washing,rinsing, pickling (3.5 cc HF 50%; 2.9 cc HNO 3 65% and 57 cc AMB),neutralizing in 0.1 M Al(NO 3 ) 3 , 9 H 2 O and ultrasonic rinsing/washing.Measurement performed are weight gain during corrosion, hardness test andmicrostructure observation using microscope optic. The results show thatβ-quenching of Zr alloy which was followed by tempering can turn αmartensite into tempered α 1 martensit. The increase of temperingtemperature decreases the Zr alloy hardness and the lowest hardness ispossessed by Zr-1% Nb alloy. The corrosion test at 275 o C and 325 o C showsthat the weight gain depends on the tempering temperature, the temperingtemperature of 400 o C and 200 o C gives the maximum weight gain for Zry-2,Zry-4 cast, Zr-1% Nb. The largest number of hydride formed during corrosionis found in Zry-2, while the small one is in Zr-1% Nb. (author)

  3. 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

  4. 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

  5. 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

  6. 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)

  7. 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

  8. 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

  9. Corrosion resistance of zirconium in nitric acid

    International Nuclear Information System (INIS)

    Kajimura, H.; Morikawa, H.; Nagano, H.

    1987-01-01

    Slow strain rate tests are effected on zirconium in boiling nitric acid to study the influence of nitric acid concentration, of oxidizing ions (Cr and Ce) and of electric potential. Corrosion resistance is excellent and stress corrosion cracking occurs only for severe conditions: 350 mV over electric potential for corrosion with nitric acid concentration of 40 % [fr

  10. Corrosion behavior of Zr-x(Nb, Sn and Cu) binary alloys

    International Nuclear Information System (INIS)

    Kim, M. H.; Lee, M. H.; Park, S. Y.; Jung, Y. H.; We, M. Y.

    1999-01-01

    For the development of advanced zirconium alloys for nuclear fuel cladding, the corrosion behaviors of zirconium binary alloys were studied on the Zr-xNb, Zr-xSn, and Zr-xCu alloys. The corrosion test were performed in water at 360 deg C, steam at 400 deg C and LiOH at 360 deg C for 45 days. The corrosion behaviors of Zr-xNb was similar to that of Zr-xCu alloys. However, the corrosion behavior of Zr-xSn was different from Zr-xNb and Zr-xCu. The weight gain of Zr-xNb and Zr-xCu was increased with addition of alloying elements. When Sn is added to Zr matrix in range below the solubility limit, the corrosion resistance decrease with increasing Sn-content, while in the range over solubility limit, Sn has an adverse effect on the corrosion resistance. Especially, Zr-xSn alloys showed higher corrosion resistance than Zr-xNb and Zr-xCu alloys in LiOH solution

  11. 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.)

  12. 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)

  13. 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

  14. 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

  15. Ternary cobalt-molybdenum-zirconium coatings for alternative energies

    Science.gov (United States)

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

    2017-11-01

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

  16. Fracture resistance of Zr–Nb alloys under low-cycle fatigue tests

    Energy Technology Data Exchange (ETDEWEB)

    Nikulin, S.A.; Rozhnov, A.B. [The National University of Science and Technology ‘‘MISIS’’, Leninsky pr. 4, 119049 Moscow (Russian Federation); Gusev, A.Yu. [A.A. Bochvar All-Russia Research Institute of Inorganic Materials (VNIINM), Rogova St. 5a, 123060 Moscow (Russian Federation); Nechaykina, T.A. [The National University of Science and Technology ‘‘MISIS’’, Leninsky pr. 4, 119049 Moscow (Russian Federation); Rogachev, S.O., E-mail: csaap@mail.ru [The National University of Science and Technology ‘‘MISIS’’, Leninsky pr. 4, 119049 Moscow (Russian Federation); Zadorozhnyy, M.Yu. [The National University of Science and Technology ‘‘MISIS’’, Leninsky pr. 4, 119049 Moscow (Russian Federation)

    2014-03-15

    Highlights: •Low-cycle fatigue tests of Zr–Nb alloys using DMA have been carried out. •The characteristics of low-cycle fatigue of the Zr–Nb alloy at 25/350 °C were determined. •Increasing test temperature up to 350 °C leads to a decrease of fatigue life. •The test temperature doesn’t have an effect on the character of fatigue curves. -- Abstract: Comparative low-cycle fatigue tests of small-scale specimens cut from the cladding tubes of E110, E125, E110opt zirconium alloys at temperatures of 25 and 350 °C using a dynamic mechanical analyzer have been carried out. It is shown that the limited cycles fatigue stress for all alloys is 50% less at temperature of 350 °C comparing to 25 °C. Besides it has been revealed that the limited cycles fatigue stress increases with increasing the strength of zirconium alloy.

  17. 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

  18. 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.

  19. 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

  20. 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

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

    Energy Technology Data Exchange (ETDEWEB)

    Das, N., E-mail: nirupamd@barc.gov.in; Sengupta, P.; Abraham, G.; Arya, A.; Kain, V.; Dey, G.K.

    2016-08-15

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

  2. 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.

  3. 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.)

  4. 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

  5. A contribution to the study of arc melting in inert gas atmospheres of zirconium sponge

    International Nuclear Information System (INIS)

    Julio Junior, O.

    1990-01-01

    Mettalic zirconium is a material of great interest in the nuclear industry due to its low thermal neutron cross section, high strength and corrosion resistance. The latter permits its use in the chemical industry. In this study, a critical bibliographic revision of the industrial processes used for the melting and consolidation of zirconium sponge has been carried out. A procedure for the melting of zirconium on a laboratory scale, has been established. An nonconsumable-electrode arc furnace have been used. The effect of process variables like atmosphere, melting current and getter, have been showed. The influence of sponge characteristics on the qualities of cast zirconium buttons have been studied. The present study is a contribution towards future investigations to obtain high purity cast zirconium and its alloys commercially known as zircaloy. (author)

  6. 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.

  7. 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

  8. 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

  9. 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)

  10. 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

  11. Performance of U-Pu-Zr fuel cast into zirconium molds

    International Nuclear Information System (INIS)

    Crawford, D.C.; Lahm, C.E.; Tsai, H.

    1992-01-01

    Current fabrication techniques for the integral fast reactor (IFR) fuel utilize injection casting into quartz molds after reprocessing in the IFR fuel cycle facility. The quartz molds are destroyed during the fuel demolding process, and the quartz residue must therefore be treated as contaminated waste. Alternatively, if the fuel can be cast into molds that remain as part of the fuel slugs (i.e., if the fuel can be left inside the molds for irradiation), then the quartz mold contribution to the waste stream can be eliminated. This possibility is being addresssed in an ongoing effort to evaluate the irradiation performance of fuel cast into zirconium sheaths rather than quartz molds. Zirconium was chosen as the sheath material because it is the component of the U-Pu-Zr fuel alloy that raises the alloy solidus temperatures and provides resistance to fuel-cladding chemical interaction (FCCI)

  12. 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)

  13. Electrochemical impedance spectroscopic study of passive zirconium

    Energy Technology Data Exchange (ETDEWEB)

    Ai Jiahe; Chen Yingzi [Center for Electrochemical Science and Technology, Department of Materials Science and Engineering, Pennsylvania State University, University Park, PA 16802 (United States); Urquidi-Macdonald, Mirna [Department of Engineering Science and Mechanics, Pennsylvania State University, University Park, PA 16802 (United States); Macdonald, Digby D. [Center for Electrochemical Science and Technology, Department of Materials Science and Engineering, Pennsylvania State University, University Park, PA 16802 (United States)], E-mail: ddm2@psu.edu

    2008-09-30

    Spent, unreproccessed nuclear fuel is generally contained within the operational fuel sheathing fabricated from a zirconium alloy (Zircaloy 2, Zircaloy 4, or Zirlo) and is then stored in a swimming pool and/or dry storage facilities until permanent disposal in a licensed repository. During this period, which begins with irradiation of the fuel in the reactor during operation, the fuel sheathing is exposed to various, aggressive environments. The objective of the present study was to characterize the nature of the passive film that forms on pure zirconium in contact with an aqueous phase [0.1 M B(OH){sub 3} + 0.001 M LiOH, pH 6.94] at elevated temperatures (in this case, 250 deg. C), prior to storage, using electrochemical impedance spectroscopy (EIS) with the data being interpreted in terms of the point defect model (PDM). The results show that the corrosion resistance of zirconium in high temperature, de-aerated aqueous solutions is dominated by the outer layer. The extracted model parameter values can be used in deterministic models for predicting the accumulation of general corrosion damage to zirconium under a wide range of conditions that might exist in some repositories.

  14. TERNARY ALLOYS OF URANIUM, COLUMBIUM, AND ZIRCONIUM

    Science.gov (United States)

    Foote, F.G.

    1960-08-01

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

  15. Reactor fuel cladding tube with excellent corrosion resistance and method of manufacturing the same

    International Nuclear Information System (INIS)

    Okuda, Takanari; Kanehara, Mitsuo; Abe, Katsuhiro; Nishimura, Takashi.

    1995-01-01

    The present invention provides a fuel cladding tube having an excellent corrosion resistance and thus a long life, and a suitable manufacturing method therefor. Namely, in the fuel cladding tube, the outer circumference of an inner layer made of a zirconium base alloy is coated with an outer layer made of a metal more corrosion resistant than the zirconium base alloy. Ti or a titanium alloy is suitable for the corrosion resistant metal. In addition, the outer layer can be coated by a method such as vapor deposition or plating, not limited to joining of the inner layer material and the outer layer material. Specifically, a composite material having an inner layer made of a zirconium alloy coated by the outer material made of a titanium alloy is applied with hot fabrication at a temperature within a range of from 500 to 850degC and at a fabrication rate of not less than 5%. The fabrication method includes any of extrusion, rolling, drawing, and casting. As the titanium-base alloy, a Ti-Al alloy or a Ti-Nb alloy containing Al of not more than 20wt%, or Nb of not more than 20wt% is preferred. (I.S.)

  16. On the mechanical effects of a nanocrystallisation treatment for ZrO2 oxide films growing on a zirconium alloy

    International Nuclear Information System (INIS)

    Panicaud, B.; Grosseau-Poussard, J.-L.; Retraint, D.; Guérain, M.; Li, L.

    2013-01-01

    Highlights: ► Raman spectroscopy is performed to determine the stress evolution in a Zr/ZrO 2 system. ► Analytical relations are used to determine material characteristics. ► A specific modelling of the mechanical fields within the oxide is done. ► Relaxation and growth parameters are identified from an inverse method. - Abstract: In the present work, mechanical features are investigated in the case of ZrO 2 thermal oxide films growing on a Zr alloy at the temperature of 550 °C. The effects of a nanocrystallisation treatment on high temperature oxidation of a zirconium alloy are specifically studied. High temperature oxidation is performed in order to show benefits of such a nanocrystallisation on corrosion resistance and its influence on the mechanical fields. Experimental results obtained by Raman spectroscopy give the growth stress evolution in ZrO 2 films. Using a modelling of the system, both asymptotic forms and an optimization procedure are developed to determine the mechanical characteristic parameters of the system.

  17. 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.

  18. 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.

  19. Orbital hybridization, crystal structure and anomalous resistivity of ultrathin CrZrx alloy films on polymeric substrates

    International Nuclear Information System (INIS)

    Evans, Drew; Zuber, Kamil; Merkens, Kerstin; Murphy, Peter

    2012-01-01

    The orbital hybridization and crystal structure are experimentally explored for ultrathin chrome zirconium (CrZr x ) alloy films co-sputtered on precoated polymeric substrates. We determine the level of orbital hybridization and crystal structure using X-ray photoelectron spectroscopy and electron diffraction. Body-centred cubic and Ω-hexagonally close-packed phases are observed to coexist in the sputtered Cr-based films. Experiments reveal the orbital hybridization and crystal structure combine to produce anomalous resistivity for these ultrathin films.

  20. Nodular Corrosion Characteristics of Zirconium Alloys

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Hyun Gil; Jeong, Y. H.; Park, S. Y.; Lee, D. J

    2003-01-15

    This study was reported the effect of the nodular corrosion on the nuclear reactor environmental along with metallurgical influence, also suggested experimental scheme related to evaluate nodular corrosion characteristics of Zr-1 Nb alloy. Remedial strategies against the nodular corrosion should firstly develop plan to assess the effect of the water quality condition (Oxygen, Hydrogen) as well as the boiling on the nodular corrosion, secondarily establish plan to control heat treatment process to keep a good resistance on nodular corrosion in Zr-1Nb alloy as former western reactor did.

  1. Aqueous corrosion study on U-Zr alloy

    International Nuclear Information System (INIS)

    Pal, Titas; Venkatesan, V.; Kumar, Pradeep; Khan, K.B.; Kumar, Arun

    2009-01-01

    In low power or research reactor, U-Zr alloy is a potential candidate for dispersion fuel. Moreover, Zirconium has a low thermal-neutron cross section and uranium alloyed with Zr has excellent corrosion resistance and dimensional stability during thermal cycling. In the present study aqueous corrosion behavior of U-Zr alloy samples was studied in autoclave at 200 deg C temperature. Corrosion rate was determined from weight loss with time. (author)

  2. 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)

  3. Crack behavior of oxidation resistant coating layer on Zircaloy-4 for accident tolerant fuel claddings

    International Nuclear Information System (INIS)

    Park, Jung Hwan; Kim, Eui Jung; Jung, Yang Il; Park, Dong Jun; Kim, Hyun Gil; Park, Jeong Yong; Yang, Jae Ho

    2016-01-01

    Terrani et al. reported the oxidation resistance of Fe-based alloys for protecting zirconium alloys from the rapid oxidation in a high-temperature steam environment. Kim and co-workers also reported the corrosion behavior of Cr coated zirconium alloy using a plasma spray and laser beam scanning. Cracks are developed by tensile stress, and this significantly deteriorates the oxidation resistance. This tensile stress is possibly generated by the thermal cycle or bending or the irradiation growth of zirconium. In this study, Cr was deposited by AIP on to Zircaloy-4 plate, and the crack behavior of Cr coated Zircaloy-4 under uni-axial tensile strain was observed. In addition, the strain of the as-deposited state was calculated by iso-inclination method. Coating began to crack at 8% of applied strain. It is assumed that a well-densified structure by AIP tends to be resistant to cracking under tensile strain.

  4. Crack behavior of oxidation resistant coating layer on Zircaloy-4 for accident tolerant fuel claddings

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jung Hwan; Kim, Eui Jung; Jung, Yang Il; Park, Dong Jun; Kim, Hyun Gil; Park, Jeong Yong; Yang, Jae Ho [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    Terrani et al. reported the oxidation resistance of Fe-based alloys for protecting zirconium alloys from the rapid oxidation in a high-temperature steam environment. Kim and co-workers also reported the corrosion behavior of Cr coated zirconium alloy using a plasma spray and laser beam scanning. Cracks are developed by tensile stress, and this significantly deteriorates the oxidation resistance. This tensile stress is possibly generated by the thermal cycle or bending or the irradiation growth of zirconium. In this study, Cr was deposited by AIP on to Zircaloy-4 plate, and the crack behavior of Cr coated Zircaloy-4 under uni-axial tensile strain was observed. In addition, the strain of the as-deposited state was calculated by iso-inclination method. Coating began to crack at 8% of applied strain. It is assumed that a well-densified structure by AIP tends to be resistant to cracking under tensile strain.

  5. Study of the oxidation behavior of zirconoium and its alloys

    International Nuclear Information System (INIS)

    Costa, I.

    1985-01-01

    The oxidation behavior of zirconium, zircaloy-4 and Zr-2,5% Nb alloy, as well as the influence of temperature, oxidising atmosphere, metal composition, heat treatment, surface treatment and specimen size on the oxidation of these materials in the temperature range 350 - 900 0 C and at atmospheric pressure have been studied with the aid of thermogravimetry. The results indicate that oxidation rate increases with temperature and the rate of oxidation of the zirconium alloys was appreciable beyond 600 0 C. At temperature higher than 500 0 C, the oxidation curves of the zirconium alloys revealed a rate transition, the kinetics after transition being either mixed parabolic and linear or linear. The transition produced an alteration in oxide characteristics, from being dark and adherent and protective, to white or grey and revealing at times cracks and scaling. The oxidation atmospheres were oxygen and air, and the results showed that the extent of oxidation in air was higher than that in oxygen. Among the metals, zirconium showed a low degree of oxidation, and the alloy Zr-2,5% Nb the lowest resistance to oxidation. Specimens heat treated in the α-phase showed the highest resistance to oxidation, and those heat treated in the β-phase the lowest. Surface treatments in aqueous solutions containing a high concentration of the fluoride ion, left behind fluorates on the surface and increased the oxidation rates of zirconium and zircaloy-4. Specimens with a high proportion of corners in relation to the total area, showed a high extent of oxidation giving rise to cracks in the oxide at the corners. (Author) [pt

  6. 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

  7. 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

  8. Relaxation resistance of heat resisting alloys with cobalt

    International Nuclear Information System (INIS)

    Borzdyka, A.M.

    1977-01-01

    Relaxation resistance of refractory nickel-chromium alloys containing 5 to 14 % cobalt is under study. The tests involve the use of circular samples at 800 deg to 850 deg C. It is shown that an alloy containing 14% cobalt possesses the best relaxation resistance exceeding that of nickel-chromium alloys without any cobalt by a factor of 1.5 to 2. The relaxation resistance of an alloy with 5% cobalt can be increased by hardening at repeated loading

  9. 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.

  10. 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.)

  11. 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

  12. 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

  13. 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

  14. 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, ...

  15. Research into zirconium alloys resistant to carbon dioxide under pressure at temperatures of up to 600 deg C (1963); Recherche d'alliages de zirconium compatibles avec le gaz carbonique sous pression jusqu'a 500 ou 600 deg C (1063)

    Energy Technology Data Exchange (ETDEWEB)

    Baque, P; Dominget, R; Bossard, J [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1963-07-01

    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{sub 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) [French] Le zirconium se trouve parmi les metaux a section de capture neutronique relativement faible et possede une temperature de fusion elevee; aussi peut on songer a l'employer notamment comme materiau de gainage d'elements combustibles pour reacteurs nucleaires refroidis au gaz carbonique. Une etude prealable de plusieurs qualites de zirconium a montre que le metal est deja assez fortement oxyde dans ce gaz des 500 deg C. En effet, le phenomene de ''breakaway'' est general; la vitesse d'oxydation devient alors lineaire et depend de la pression du gaz carbonique. La recherche d'alliages binaires et ternaires a donc ete entreprise afin de tenter d'ameliorer le comportement du metal. Elle a permis d'aboutir a quelques compositions interessantes: cuivre 1, 1,6 et 2,5 pour cent, vanadium 2 pour cent, et calcium 0,05 et 0,5 pour cent. Des alliages ternaires au cuivre-molybdene et cuivre-phosphore sont egalement moins oxydables, et en particulier ne presentent pas le phenomene de ''breakaway'', meme apres une longue exposition a 600 deg C. (auteurs)

  16. Analyses on Silicide Coating for LOCA Resistant Cladding

    Energy Technology Data Exchange (ETDEWEB)

    Sweidan, Faris B.; Lee, You Ho; Ryu, Ho Jin [KAIST, Daejeon (Korea, Republic of)

    2015-10-15

    A particular focus of accident-tolerant fuel has been cladding due to the rapid high-temperature oxidation of zirconium-based cladding with the evolution of H2 when steam is a reactant. Some key features of the coated cladding include high-temperature resistance to oxidation, lower processing temperatures, and a high melting point of the coating. Zirconium alloys exhibit a reasonably high melting temperature, so a coating for the cladding is appealing if the coating increases the high-temperature resistance to oxidation. In this case, the cladding is protected from complete oxidation. The cladding coating involves the application of zirconium silicide onto Zr-based cladding. Zirconium silicide coating is expected to produce a glassy layer that becomes more protective at elevated temperature. For this reason, silicide coatings on cladding offer the potential for improved reliability at normal operating temperatures and at the higher transient temperatures encountered during accidents. Although ceramic coatings are brittle and may have weak points to be used as coating materials, several ceramic coatings were successful and showed adherent behavior and high resistance to oxidation. In this study, the oxidation behavior of zirconium silicide and its oxidation kinetics are analyzed. Zirconium silicide is a new suggested material to be used as coatings on existing Zr-based cladding alloys, the aim of this study is to evaluate if zirconium silicide is applicable to be used, so they can be more rapidly developed using existing cladding technology with some modifications. These silicide coatings are an attractive alternative to the use of coatings on zirconium claddings or to the lengthy development of monolithic ceramic or ceramic composite claddings and coatings.

  17. Analyses on Silicide Coating for LOCA Resistant Cladding

    International Nuclear Information System (INIS)

    Sweidan, Faris B.; Lee, You Ho; Ryu, Ho Jin

    2015-01-01

    A particular focus of accident-tolerant fuel has been cladding due to the rapid high-temperature oxidation of zirconium-based cladding with the evolution of H2 when steam is a reactant. Some key features of the coated cladding include high-temperature resistance to oxidation, lower processing temperatures, and a high melting point of the coating. Zirconium alloys exhibit a reasonably high melting temperature, so a coating for the cladding is appealing if the coating increases the high-temperature resistance to oxidation. In this case, the cladding is protected from complete oxidation. The cladding coating involves the application of zirconium silicide onto Zr-based cladding. Zirconium silicide coating is expected to produce a glassy layer that becomes more protective at elevated temperature. For this reason, silicide coatings on cladding offer the potential for improved reliability at normal operating temperatures and at the higher transient temperatures encountered during accidents. Although ceramic coatings are brittle and may have weak points to be used as coating materials, several ceramic coatings were successful and showed adherent behavior and high resistance to oxidation. In this study, the oxidation behavior of zirconium silicide and its oxidation kinetics are analyzed. Zirconium silicide is a new suggested material to be used as coatings on existing Zr-based cladding alloys, the aim of this study is to evaluate if zirconium silicide is applicable to be used, so they can be more rapidly developed using existing cladding technology with some modifications. These silicide coatings are an attractive alternative to the use of coatings on zirconium claddings or to the lengthy development of monolithic ceramic or ceramic composite claddings and coatings

  18. 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.

  19. Development of Silicide Coating on Molybdenum Alloy Cladding

    International Nuclear Information System (INIS)

    Lim, Woojin; Ryu, Ho Jin

    2015-01-01

    The molybdenum alloy is considered as one of the accident tolerant fuel (ATF) cladding materials due to its high temperature mechanical properties. However, molybdenum has a weak oxidation resistance at elevated temperatures. To modify the oxidation resistance of molybdenum cladding, silicide coating on the cladding is considered. Molybdenum silicide layers are oxidized to SiO 2 in an oxidation atmosphere. The SiO 2 protective layer isolates the substrate from the oxidizing atmosphere. Pack cementation deposition technique is widely adopted for silicide coating for molybdenum alloys due to its simple procedure, homogeneous coating quality and chemical compatibility. In this study, the pack cementation method was conducted to develop molybdenum silicide layers on molybdenum alloys. It was found that the Mo 3 Si layer was deposited on substrate instead of MoSi 2 because of short holding time. It means that through the extension of holding time, MoSi 2 layer can be formed on molybdenum substrate to enhance the oxidation resistance of molybdenum. The accident tolerant fuel (ATF) concept is to delay the process following an accident by reducing the oxidation rate at high temperatures and to delay swelling and rupture of fuel claddings. The current research for Atf can be categorized into three groups: First, modification of existing zirconium-based alloy cladding by improving the high temperature oxidation resistance and strength. Second, replacing Zirconium based alloys with alternative metallic materials such as refractory elements with high temperature oxidation resistance and strength. Third, designing alternative fuel structures using ceramic and composite systems

  20. 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)

  1. Modeling of High Temperature Oxidation Behavior of FeCrAl Alloy by using Artificial Neural Network

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jae Joon; Ryu, Ho Jin [KAIST, Daejeon (Korea, Republic of)

    2016-10-15

    Refractory alloys are candidate materials for replacing current zirconium-base cladding of light water reactors and they retain significant creep resistance and mechanical strength at high temperatures up to 1500 ℃ due to their high melting temperature. Thermal neutron cross sections of refractory metals are higher than that of zirconium, however the loss of neutron can be overcome by reducing cladding thickness which can be facilitated with enhanced mechanical properties. However, most refractory metals show the poor oxidation resistance at a high temperature. Oxidation behaviors of the various compositions of FeCrAl alloys in high temperature conditions were modeled by using Bayesian neural network. The automatic relevance determination (ARD) technique represented the influence of the composition of alloying elements on the oxidation resistance of FeCrAl alloys. This model can be utilized to understand the tendency of oxidation behavior along the composition of each element and prove the applicability of neural network modeling for the development of new cladding material of light water reactors.

  2. Non-ferrous metals, anorganic and organic materials resistent to fluorides

    International Nuclear Information System (INIS)

    Hauffe, K.

    1986-01-01

    Aluminium and its alloys are resistant in fluoride solutions up to 400 K. Aluminium is also a suitable reactor material for the thermal decomposition of acidic fluorides between 750 and 825 K. Brass corrodes at room temperature in a 0,1 m KF solution with and without inhibitors very slowly ( -1 ). Nickel and the nickel alloys Inconel 600, Hastelloy N and Monel 500 are the most resistant materials against fluoride solutions and melts. A similar behavior exhibit zirconium-titanium-iron and zirconium-titanium-molybdenum alloys, respectively. From the inorganic compounds, compressed graphite, Al 2 O 3 and hexaborides of earth and rare earth metals, particularly LaB 6 , are extraordinarily resistant against fluorine ions at high temperatures. If the reaction temperature remains below 370 K, then polymers and resins, e.g. polyolefines, PVC, acrylic and epoxy resins and fluorcarbon resins can be employed as coating or compound material (resin + carbon fibers) resistant against fluorine ions up to 370 K. (orig.) [de

  3. 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

  4. 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)

  5. Manufacturing method of zirconium alloy-type structural material in reactor core excellent in corrosion resistance, especially in uniform corrosion resistance and hydrogen absorption resistance

    International Nuclear Information System (INIS)

    Mozumi, Yasuhiro.

    1997-01-01

    A zirconium alloy comprising from 0.8 to 1.6wt% of Sn, from 0.17 to 0.25wt% of Fe, from 0.15 to 0.25wt% of Cr and from 0.01 to 0.08wt% of Ni and Si at a concentration of 120ppm or lower as an impurity and the balance of Zr is melted into cast pieces and then subjected to an β annealing. It is controlled so as to satisfy Fe + Cr + Ni ≤ 0.52wt%. Then, rolling and annealing are applied so that the total heat injection amount ΣA i to the materials is within a range of from 1 x 10 -19 to 1 x 10 -17 . ΣA i = Σt i · exp(-Q/RT i ), in which t i represents processing time (hour) at an ith heat treatment step after the β annealing, T i represents a processing temperature (K) in the step i. Q represents an activating energy, R represents a gas constant, and Q/R 40,000. (I.N.)

  6. High-temperature air oxidation of E110 and Zr-1%Nb alloys claddings with coatings

    International Nuclear Information System (INIS)

    Kuprin, A.S.; Belous, V.A.; Voyevodin, V.N.; Bryk, V.V.; Vasilenko, R.L.; Ovcharenko, V.D.; Tolmachova, G.N.; V'yugov, P.N.

    2014-01-01

    Results of experimental study of the influence of protective vacuum-arc claddings on the base of compounds zirconium-chromium and of its nitrides on air oxidation resistance at temperatures 660, 770, 900, 1020, 1100 deg C during 3600 s. of tubes produced of zirconium alloys E110 and Zr-1%Nb (calcium-thermal alloy of Ukrainian production) are presented. Change of hardness, the width of oxide layer and depth of oxygen penetration into alloys from the side of coating and without coating are investigated by the methods of nanoindentation and by scanning electron microscopy. It is shown that the thickness of oxide layer in zirconium alloys at temperatures 1020 and 1100 deg C from the side of the coating doesn't exceed 5 μm, and from the unprotected side reaches the value of ≥ 120 μm with porous and rough structure. Tubes with coatings save their shape completely independently of the type of alloy; tubes without coatings deform with the production of through cracks

  7. 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

  8. 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

  9. 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)

  10. 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.

  11. 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

  12. 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

  13. 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)

  14. 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)

  15. 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

  16. 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)

  17. 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

  18. 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.

  19. 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.

  20. Evaluation of corrosion resistance of implant-use Ti-Zr binary alloys with a range of compositions.

    Science.gov (United States)

    Akimoto, Teisuke; Ueno, Takeshi; Tsutsumi, Yusuke; Doi, Hisashi; Hanawa, Takao; Wakabayashi, Noriyuki

    2018-01-01

    Although titanium-zirconium (Ti-Zr) alloy has been adopted for clinical applications, the ideal proportion of Zr in the alloy has not been identified. In this study, we investigated the biocompatibility of Ti-Zr alloy by evaluating its corrosion resistance to better understand whether there is an optimal range or value of Zr proportion in the alloy. We prepared pure Ti, Ti-30Zr, Ti-50Zr, Ti-70Zr, and pure Zr (mol% of Zr) samples and subjected them to anodic polarization and immersion tests in a lactic acid + sodium chloride (NaCl) solution and artificial saliva. We observed pitting corrosion in the Ti-70Zr and Zr after exposure to both solutions. After the immersion test, we found that pure Ti exhibited the greatest degree of dissolution in the lactic acid + NaCl solution, with the addition of Zr dramatically reducing Ti ion dissolution, with the reduction ultimately exceeding 90% in the case of the Ti-30Zr. Hence, although the localized corrosion resistance under severe conditions was compromised when the Zr content was more than 70%, metal ion release reduced owing to Zr addition and the corresponding formation of a stable passive layer. The results suggest that Ti-30Zr or a Zr proportion of less than 50% would offer an ideal level of corrosion resistance for clinical applications. © 2016 Wiley Periodicals, Inc. J Biomed Mater Res Part B: Appl Biomater, 106B: 73-79, 2018. © 2016 Wiley Periodicals, Inc.

  1. 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)

  2. Strength and gas-abrasive wear-resistance of zirconium carbide based cerments

    International Nuclear Information System (INIS)

    Samsonov, G.V.; Dan'kin, A.A.; Markov, A.A.; Bogomol, I.V.

    1976-01-01

    Results relating to a study of cermet strength and wear resistance by means of a gas-abrasive flow are presented. It has been found that with a higher amount of the metallic binder (over 25 at.%) in zirconium carbide-based cermets the bending and compression strength and also hardness and wear resistance within the systems ZrC-Nb, ZrC-Mo, ZrC-W become lower. The interrelation of the cermet wear resistance of the various systems and their bending and compression strengths, which, in turn, depend on the electronic structure is shown

  3. The corrosion behaviour of Zr3Al-based alloys

    International Nuclear Information System (INIS)

    Murphy, E.V.; Wieler, R.

    1977-07-01

    The corrosion resistance of several zirconium-aluminum alloys with aluminum contents ranging from 7.6 to 9.6 wt% was examined in 300 deg C and 325 deg C water, 350 deg C and 400 deg C steam and in air and wet CO 2 at 325 deg C and 400 deg C. In the transformed alloys there are three phases present, αZr, Zr 2 Al and Zr 3 Al of which the αZr phase is the least corrosion resistant. The most important factor controlling the corrosion behaviour of these alloys was found to be the size, distribution and amount of the αZr phase in the transformed alloys, which in turn was dependent upon the microstructural scale of the untransformed alloys

  4. 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.

  5. 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)

  6. Alloying principles for magnesium base heat resisting alloys

    International Nuclear Information System (INIS)

    Drits, M.E.; Rokhlin, L.L.; Oreshkina, A.A.; Nikitina, N.I.

    1982-01-01

    Some binary systems of magnesium-base alloys in which solid solutions are formed, are considered for prospecting heat resistant alloys. It is shown that elements having essential solubility in solid magnesium strongly decreasing with temperature should be used for alloying maqnesium base alloys with high strength properties at increased temperatures. The strengthening phases in these alloys should comprise essential quantity of magnesium and be rather refractory

  7. 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

  8. Zirconium-barrier cladding attributes

    International Nuclear Information System (INIS)

    Rosenbaum, H.S.; Rand, R.A.; Tucker, R.P.; Cheng, B.; Adamson, R.B.; Davies, J.H.; Armijo, J.S.; Wisner, S.B.

    1987-01-01

    This metallurgical study of Zr-barrier fuel cladding evaluates the importance of three salient attributes: (1) metallurgical bond between the zirconium liner and the Zircaloy substrate, (2) liner thickness (roughly 10% of the total cladding wall), and (3) softness (purity). The effect that each of these attributes has on the pellet-cladding interaction (PCI) resistance of the Zr-barrier fuel was studied by a combination of analytical model calculations and laboratory experiments using an expanding mandrel technique. Each of the attributes is shown to contribute to PCI resistance. The effect of the zirconium liner on fuel behavior during off-normal events in which steam comes in contact with the zirconium surface was studied experimentally. Simulations of loss-of-coolant accident (LOCA) showed that the behavior of Zr-barrier cladding is virtually indistinguishable from that of conventional Zircaloy cladding. If steam contacts the zirconium liner surface through a cladding perforation and the fuel rod is operated under normal power conditions, the zirconium liner is oxidized more rapidly than is Zircaloy, but the oxidation rate returns to the rate of Zircaloy oxidation when the oxide phase reaches the zirconium-Zircaloy metallurgical bond

  9. Crystallization characteristics of amorphous alloys of FeZr

    International Nuclear Information System (INIS)

    Rozhan, M. Idrus; Grundy, P.J.

    1993-01-01

    The crystallization characteristics of sputter-deposited amorphous alloys of Fe 100-x Zr x prepared at zirconium concentrations between 9 and 89 at.% was investigated. The transformation of the alloys from the amorphous to the crystalline state has been examined by thermal analysis, electrical resistance and X-ray diffraction. The crystallization temperatures were determined by differential scanning calorimetry (DSC) and electrical resistance as a function of temperature. The final phases were determined by X-ray diffraction. The activation energies were calculated from the Kissinger plots and the heats of crystallization were calculated and correlations between the thermal analysis and the resistance results are presented

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

    Science.gov (United States)

    Rebak, Raul B.

    2017-12-01

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

  11. 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)

  12. 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)

  13. 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.

  14. 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.

  15. 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

  16. Electrochemical Impedance Spectroscopy Of Metal Alloys

    Science.gov (United States)

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

    1993-01-01

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

  17. Zirconium for nitric acid solutions

    International Nuclear Information System (INIS)

    Yau, T.L.

    1984-01-01

    The excellent corrosion resistance of zirconium in nitric acid has been known for over 30 years. Recently, there is an increasing interest in using zirconium for nitric acid services. Therefore, an extensive research effort has been carried out to achieve a better understanding of the corrosion properties of zirconium in nitric acid. Particular attention is paid to the effect of concentration, temperature, structure, solution impurities, and stress. Immersion, autoclave, U-bend, and constant strain-rate tests were used in this study. Results of this study indicate that the corrosion resistance of zirconium in nitric acid is little affected by changes in temperature and concentration, and the presence of common impurities such as seawater, sodium chloride, ferric chloride, iron, and stainless steel. Moreover, the presence of seawater, sodium chloride, ferric chloride, and stainless steel has little effect on the stress corrosion craking (SCC) susceptibility of zirconium in 70% nitric acid at room temperatures. However, zirconium could be attacked by fluoride-containing nitric acid and the vapors of chloride-containing nitric acid. Also, high sustained tensile stresses should be avoided when zirconium is used to handle 70% nitric acid at elevated temperatures or > 70% nitric acid

  18. Estimation of weld nugget temperature by thermography method in resistance projection welding process

    International Nuclear Information System (INIS)

    Setty, D.S.; Rameswara Roa, A.; Hemantha Rao, G.V.S.; Jaya Raj, R.N.

    2008-01-01

    In the Pressurized Heavy Water Reactor (PHWR) fuel manufacturing, zirconium alloy appendages like spacer and bearing pads are welded to the thin wall zirconium alloy fuel tubes by using resistance projection welding process. Out of many joining processes available, resistance-welding process is reliable, environment friendly and best suitable for mass production applications. In the fuel assembly, spacer pads are used to get the required inter-element spacing and Bearing pads are used to get the required load-bearing surface for the fuel assembly. Performance of the fuel assembly in the reactor is greatly influenced by these weld joint's quality. Phase transformation from α to β phase is not acceptable while welding these tiny appendages. At present only destructive metallography test is available for this purpose. This can also be achieved by measuring weld nugget temperature where in the phase transformation temperature for zirconium alloy material is 853 o C. The temperature distribution during resistance welding of tiny parts cannot be measured by conventional methods due to very small space and short weld times involved in the process. Shear strength, dimensional accuracy and weld microstructures are some of the key parameters used to measure the quality of appendage weld joints. Weld parameters were optimized with the help of industrial experimentation methodology. Individual projection welding by split electrode concept, and during welding on empty tube firm support is achieved on inner side of the tube by using expandable pneumatic mandrel. In the present paper, an attempt was made to measure the weld nugget temperature by thermography technique and is correlated with standard microstructures of zirconium alloy material. The temperature profiles in the welding process are presented for different welding conditions. This technique has helped in measuring the weld nugget temperature more accurately. It was observed that in the present appendage welding

  19. Effect of sintering processing on microstructure, mechanical properties and corrosion resistance of Ti–24Nb–4Zr–7.9Sn alloy for biomedical applications

    International Nuclear Information System (INIS)

    Guo, Shibo; Chu, Aimin; Wu, Haijiang; Cai, Chunbo; Qu, Xuanhui

    2014-01-01

    Highlights: • Ti–24Nb–4Zr–7.9Sn alloy is prepared by powder metallurgy method. • The alloy prepared at 1250 °C for 2 h has more β-matrix and tiny α-precipitation. • The alloy prepared at 1250 °C for 2 h possesses good mechanical properties. • The alloy prepared at 1250 °C for 2 h exhibits better corrosion resistance. - Abstract: Ti–24Nb–4Zr–7.9Sn alloy was prepared by Powder Metallurgy (PM) method using titanium hydride powder, niobium powder, zirconium powder, and tin powder as raw materials. The effect of sintering processing on microstructure, mechanical properties, and corrosion resistance was investigated in details. The alloy possessed dominant β-matrix and a little α-precipitation. The mechanical properties of the alloy sintered at 1250 °C for 2 h were better than those of the alloys with other sintering processing, which would avoid stress shielding and thus prevent bone resorption in orthopedic implants applications. As long-term stability in biological environment is required, the electrochemical behaviors in a simulated body fluid (Hank’s solution and simulated saliva solution) were also evaluated. Potentiodynamic polarization curves exhibited that the sample sintered at 1250 °C for 2 h had better corrosion properties than those of other sintering processing. The good corrosion resistance combined with better mechanical biocompatibility made the Ti–24Nb–4Zr–7.9Sn alloy suitable for use as orthopedic implants

  20. Evaluation of delayed hydride cracking and fracture toughness in zirconium alloys

    International Nuclear Information System (INIS)

    Oh, Je Yong

    2000-02-01

    The tensile, fracture toughness, and delayed hydride cracking (DHC) test were carried at various temperatures to understand the effect of hydrides on zirconium alloys. And the effects of yield stress and texture on the DHC velocity were discussed. The tensile properties of alloy A were the highest, and the difference between directions in alloy C was small due to texture. The fracture toughness at room temperature decreased sharply when hydrided. Although the alignment of hydride plates was parallel to loading direction, the hydrides were fractured due to the triaxiality at the crack tip region. The fracture toughness over 200 .deg. C was similar regardless of the hydride existence, because the triaxiality region was lost due to the decrease of yield stress with temperature. As the yield stress decreased, the threshold stress intensity factor and the striation spacing increased in alloy A, and the fracture surfaces and striations were affected by microstructures in all alloys. To evaluate the effect of the yield stress on DHC velocity, a normalization method was proposed. When the DHC velocity was normalized with dividing by the terminal solid solubility and the diffusion coefficient of hydrogen, the relationship between the yield stress and the DHC velocity was representable on one master curve. The equation from the master curve was able to explain the difference between the theoretical activation energy and the experimental activation energy in DHC. The difference was found to be ascribed to the decrease of yield stress with temperature. texture affected the delayed hydride cracking velocity by yield stress and by hydride reprecipitation. The relationship between the yield stress and the DHC velocity was expressed as an exponential function, and the relationship between the reprecipitation of hydride and the DHC velocity was expressed as a linear function

  1. Viscoplastic behavior of zirconium alloys in the temperatures range 20 deg C - 400 deg C: characterization and modeling of strain ageing phenomena

    International Nuclear Information System (INIS)

    Graff, St.

    2006-10-01

    The anomalous strain rate sensitivity of zirconium alloys over the temperatures range 20-600 C has been widely reported in the literature. This unconventional behavior is related to the existence of strain ageing phenomenon which results from the combined action of thermally activated diffusion of foreign atoms to and along dislocation cores and the long range of dislocations interactions. The important role of interstitial and substitutional atoms in zirconium alloys, responsible for strain ageing and the lack of information about the domain where strain ageing is active have not been yet adequately characterized because of the multiplicity of alloying elements and chemical impurities. The aim of this work is to characterize experimentally the range of temperatures and strain rates where strain ageing is active on the macroscopic and mesoscopic scales. We propose also a predictive approach of the strain ageing effects, using the macroscopic strain ageing model suggested by McCormick (McCormick, 1988; Zhang et al., 2000). Specific zirconium alloys were elaborated starting from a crystal bar of zirconium with 2.2 wt% hafnium and very low oxygen content (80 wt ppm), called ZrHf. Another substitutional atom was added to the solid solution under the form of 1 wt% niobium. Some zirconium alloys were doped with oxygen, others were not. All of them were characterized by various mechanical tests (standard tensile tests, tensile tests with strain rate changes, relaxation tests with unloading). The experimental results were compared with those for the standard oxygen doped zirconium alloy (1300 wt ppm) studied by Pujol (Pujol, 1994) and called Zr702. The following experimental evidences of the age-hardening phenomena were collected and then modeled: 1) low and/or negative strain rate sensitivity around 200-300 C, 2) creep arrest at 200 C, 3) relaxation arrest at 200 C and 300 C, 4) plastic strain heterogeneities observed in laser extensometry on the millimeter scale

  2. Viscoplastic behavior of zirconium alloys in the temperatures range 20 deg C - 400 deg C: characterization and modeling of strain ageing phenomena

    Energy Technology Data Exchange (ETDEWEB)

    Graff, St

    2006-10-15

    The anomalous strain rate sensitivity of zirconium alloys over the temperatures range 20-600 C has been widely reported in the literature. This unconventional behavior is related to the existence of strain ageing phenomenon which results from the combined action of thermally activated diffusion of foreign atoms to and along dislocation cores and the long range of dislocations interactions. The important role of interstitial and substitutional atoms in zirconium alloys, responsible for strain ageing and the lack of information about the domain where strain ageing is active have not been yet adequately characterized because of the multiplicity of alloying elements and chemical impurities. The aim of this work is to characterize experimentally the range of temperatures and strain rates where strain ageing is active on the macroscopic and mesoscopic scales. We propose also a predictive approach of the strain ageing effects, using the macroscopic strain ageing model suggested by McCormick (McCormick, 1988; Zhang et al., 2000). Specific zirconium alloys were elaborated starting from a crystal bar of zirconium with 2.2 wt% hafnium and very low oxygen content (80 wt ppm), called ZrHf. Another substitutional atom was added to the solid solution under the form of 1 wt% niobium. Some zirconium alloys were doped with oxygen, others were not. All of them were characterized by various mechanical tests (standard tensile tests, tensile tests with strain rate changes, relaxation tests with unloading). The experimental results were compared with those for the standard oxygen doped zirconium alloy (1300 wt ppm) studied by Pujol (Pujol, 1994) and called Zr702. The following experimental evidences of the age-hardening phenomena were collected and then modeled: 1) low and/or negative strain rate sensitivity around 200-300 C, 2) creep arrest at 200 C, 3) relaxation arrest at 200 C and 300 C, 4) plastic strain heterogeneities observed in laser extensometry on the millimeter scale

  3. Synthesis of zirconium by zirconium tetrachloride reduction by magnesio-thermia. Experimental study and modelling; Elaboration de zirconium par reduction de tetrachlorure de zirconium par magnesothermie. Etude experimentale et modelisation

    Energy Technology Data Exchange (ETDEWEB)

    Basin, N

    2001-01-01

    This work deals with the synthesis of zirconium. The ore is carbo-chlorinated to obtain the tetrachloride which is then purified by selective condensation and extractive distillation. Zirconium tetrachloride is then reduced by magnesium and the pseudo-alloy is obtained according to the global following reaction (Kroll process): ZrCl{sub 4} + 2 Mg = 2 MgCl{sub 2}. By thermodynamics, it has been shown that the volatilization of magnesium chloride and the formation of zirconium sub-chlorides are minimized when the combined effects of temperature and of dilution with argon are limited. With these conditions, the products, essentially zirconium and magnesium chloride, are obtained in equivalence ratio in the magnesio-thermia reaction. The global kinetics of the reduction process has been studied by a thermal gravimetric method. A thermo-balance device has been developed specially for this kinetics study. It runs under a controlled atmosphere and is coupled to a vapor tetrachloride feed unit. The transformation is modelled supposing that the zirconium and magnesium chloride formation result: 1)of the evaporation of magnesium from its liquid phase 2)of the transfer of magnesium and zirconium tetrachloride vapors towards the front of the reaction located in the gaseous phase 3)of the chemical reaction. In the studied conditions, the diffusion is supposed to be the limiting process. The influence of the following parameters: geometry of the reactive zone, temperature, scanning rate of the argon-zirconium tetrachloride mixture, composition of the argon-zirconium tetrachloride mixture has been experimentally studied and confronted with success to the model. (O.M.)

  4. Study for the chlorination of zirconium oxide

    International Nuclear Information System (INIS)

    Seo, E.S.M.; Takiishi, H.; Paschoal, J.O.A.; Andreoli, M.

    1990-12-01

    In the development of new ceramic and metallic materials the chlorination process constitutes step in the formation of several intermediate compounds, such as metallic chlorides, used for the production of high, purity raw materials. Chlorination studies with the aim of fabrication special zirconium-base alloys have been carried out at IPEN. Within this program the chlorination technique has been used for zirconium tetrachloride production from zirconium oxide. In this paper some relevant parameters such as: time and temperature of reaction, flow rate of chloride gas and percentage of the reducing agent which influence the efficiency of chlorination of zirconium oxide are evaluated. Thermodynamical aspects about the reactions involved in the process are also presented. (author)

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

    Science.gov (United States)

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

    2016-03-23

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

  6. 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.

  7. The Evaluation of the Corrosion Resistance of the Al-Si Alloys Antimony Alloyed

    Directory of Open Access Journals (Sweden)

    Svobodova J.

    2014-06-01

    Full Text Available This paper deals with the evaluation of the corrosion resistance of the Al-Si alloys alloyed with the different amount of antimony. Specifically it goes about the alloy AlSi7Mg0,3 which is antimony alloyed in the concentrations 0; 0,001; 0,005; 0,01 a 0,05 wt. % of antimony. The introduction of the paper is dedicated to the theory of the aluminium alloys corrosion resistance, testing and evaluation of the corrosion resistance. The influence of the antimony to the Al-Si alloys properties is described further in the introduction. The experimental part describes the experimental samples which were prepared for the experiment and further they were exposed to the loading in the atmospheric conditions for a period of the 3 months. The experimental samples were evaluated macroscopically and microscopically. The results of the experiment were documented and the conclusions in terms of the antimony impact to the corrosion resistance of the Al-Si alloy were concluded. There was compared the corrosion resistance of the Al-Si alloy antimony alloyed (with the different antimony content with the results of the Al-Si alloy without the alloying after the corrosion load in the atmospheric conditions in the experiment.

  8. Study of corrosion kinetics of fuel element tubes from calcium-thermal zirconium alloy Zr1Nb in water at 350 degree C and in vapour at 400 and 500 degree C

    International Nuclear Information System (INIS)

    Petel'guzov, I.A.

    2002-01-01

    In the report brought results of corrosion process studies in water medium of pipe samples for fuel element shells from Zr1Nb alloy (earlier KTZ-110),made from the calcium-thermal zirconium alloys developed in the Ukraine of technology and,for the comparison,samples of pipes from the staff alloy E110, applicable in fuel elements acting reactors of type WWER. Tests were conducted under the working temperature of fuel shells in the reactor (350 degree C) in during of 14000 hours and under increased temperatures (400 degree C) within a time acordinly 4000 hours. Samples from the alloy Zr1Nb had more high contents of oxygen (before 0,12%...0,16%), than staff alloy Eh110 (0,08%O). Studies have shown sufficiently high corrosion stability of experimental alloy Zr1Nb, close to stability of alloy E110.Discovered signs of corrosion 'breakway' or 'transition' on kinetic corrosion curves of Zr1Nb alloys and E110 alloy, characterisating zircaloy type of alloy. Considered mechanism of influence of oxygen on the corrosion process of zirconium alloys with the additive a niobium

  9. Corrosion resistance of titanium alloys for dentistry

    International Nuclear Information System (INIS)

    Laskawiec, J.; Michalik, R.

    2001-01-01

    Titanium and its alloys belong to biomaterials which the application scope in medicine increases. Some properties of the alloys, such as high mechanical strength, low density, low Young's modulus, high corrosion resistance and good biotolerance decide about it. The main areas of the application of titanium and its alloys are: orthopedics and traumatology, cardiosurgery, faciomaxillary surgery and dentistry. The results of investigations concerning the corrosion resistance of the technical titanium and Ti6Al14V alloy and comparatively a cobalt alloy of the Vitallium type in the artificial saliva is presented in the work. Significantly better corrosion resistance of titanium and the Ti6Al14V than the Co-Cr-Mo alloy was found. (author)

  10. Evaluation of a Zirconium Recycle Scrubber System

    Energy Technology Data Exchange (ETDEWEB)

    Spencer, Barry B. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Bruffey, Stephanie H. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2017-04-01

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

  11. Modification of structural phase state in superficial layers of fuel tubes made of Zirconium alloys

    International Nuclear Information System (INIS)

    Volkov, N.; Kalin, B.; Pimenov, Y.; Timoshin, S.

    2011-01-01

    The paper presents the results obtained in developing the method for introduction of the required changes into states and properties of outer surface on fuel rod cladding made of zirconium alloys E110 and E635 through irradiation by radial Ar + ion beam with a broad energy spectrum. In particular, the paper demonstrates that ion beam treatment of the claddings surface, at the final stage of their fabrication, can upgrade substantially quality of outer tubular surface after mechanical polishing (the cleaner surface, the lower roughness, removal of technological transversal scratches). In addition, the ion beam irradiation results in higher micro-hardness of the modified layer and in better tribological parameters. Kinetic effects in growth of oxide films were studied for the tubular samples of zirconium alloys after ion-beam treatment (cleaning and polishing by radial Ar + ion beam). Also, corrosion tests of the tubular samples were carried out in water (at 350 0 C) and steam (at 350, 375 and 400 0 C) with duration up to 3000 hours. It was revealed that oxide layer consisting mainly of zirconium dioxide in monoclinic modification was formed on tubular surface after oxidation at 3500 0 C in water or steam. The oxidizing process in the pressurized steam created thicker oxide layer on tubular surface than that in the pressurized water. Experimental data were used to determine optimal conditions for ion-beam treatment of outer fuel tube surface. The tubular samples with the following geometrical parameters were investigated: length - up to 500 mm, diameter - 9,15 mm. Optimal regimes for ion-beam cleaning and polishing of the tubular samples were studied up to the process rate of 1 meter per minute. Within the frames of linear approximation, analytical relationships were derived for time dependent growth of oxide films and used to evaluate thickness of oxide film under test conditions (duration . up to 10000 hours). Thickness of oxide films can cover the range from 6

  12. 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

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

    Science.gov (United States)

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

    2013-01-01

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

  14. Capture of Tritium Released from Cladding in the Zirconium Recycle Process

    Energy Technology Data Exchange (ETDEWEB)

    Spencer, Barry B [ORNL; Bruffey, Stephanie H [ORNL; DelCul, Guillermo Daniel [ORNL; Walker, Trenton Baird [ORNL

    2016-08-31

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

  15. Capture of Tritium Released from Cladding in the Zirconium Recycle Process

    Energy Technology Data Exchange (ETDEWEB)

    Bruffey, Stephanie H [ORNL; Spencer, Barry B [ORNL; DelCul, Guillermo Daniel [ORNL

    2016-08-31

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

  16. The effects of radiation on aluminium alloys in the core of energy nuclear reactors

    International Nuclear Information System (INIS)

    Petrossian, V.G.

    1995-01-01

    One of the attractive directions in the worldwide practice of nuclear installations is the replacement of expensive zirconium alloy with more cheap materials, particularly aluminium allo. For Heat Supply Nuclear Plants (HSNP) with approximately 473 K core temperatures, the use of heat-resistant aluminium alloys seems to be reasonable. The present work is concerned with the studies on radiation effects on aluminium alloy, and interaction between the alloy and coolant in the reactor core. (author). 2 refs., 3 figs., 1 tab

  17. Application of thermo-electromotive force and electric resistance measuring methods to researching phase transformations in Zr1Nb alloy

    International Nuclear Information System (INIS)

    Klimenko, S.P.; Gritsina, V.M.; Petel'guzov, I.A.; Chernyaeva, T.P.

    2007-01-01

    The paper determines the applicability areas of different methods for the study of structural phase transformations in a Zr+1Nb alloy, which is extensively used in reactor construction; production and fabrication of products from Zr+1Nb is currently developed in Ukraine. Electromotive force and electric resistance were measured to study structural phase transformations of Zr+1Nb fuel rod tubes based on calciumthermal zirconium (Zr1Nb). It was established that changes in electric resistance clearly show the beginning of a massive α → β transition at ∼ 750 degree C and the end of α → β transition at ∼ 950 degree C, whereas measurement of thermo-e.m.f. in the samples subjected to successive 3-hour step annealing in the temperature range from 300 to 700 degree C allows the temperature of monotectoid transformation to be found. For sample Zr1Nb batches the temperature of monotectoid transformation is (620±7) degree C. The measurement results are consistent with the similar studies carried out on Zr+1Nb fuel rod tubes based on electrolytic zirconium (E110), for which the temperature of monotectoid transformation is equal to ∼ 610 degree C

  18. Influence of alkali metal hydroxides on corrosion of Zr-based alloys

    International Nuclear Information System (INIS)

    Jeong, Y.H.; Ruhmann, H.; Garzarolli, F.

    1997-01-01

    In this study the influence of group-1 alkali hydroxides on different zirconium based alloys has been evaluated. The experiments have been carried out in small stainless steel autoclaves at 350 deg. C in pressurized 17 MPa water, with in low (0.32 mmol), medium (4.3 mmol) and high (31.5 mmol) equimolar concentrations of Li-, Na-, K-, Rb- and Cs-Hydroxides. Two types of alloys have been investigated: Zr-Sn-(Transition metal) and Zr-Sn-Nb-(Transition metal). The corrosion behaviour was evaluated from weight gain measurements. From the experiments the cation could be identified as the responsible species for zirconium alloy corrosion in alkalized water. The radius of the cation governs the corrosion behaviour in the pre accelerated region of zircaloy corrosion. Incorporating of alkali cations into the zirconium oxide lattice is probably the mechanism which allows the corrosion enhancement for Li and Na and the significantly lower effect for the other bases. Nb containing alloys show lower corrosion resistance than alloys from the Zr-Sn-TRM system in all alkali solutions. Both types of alloys corrode significantly more in LiOH and NaOH than in the other alkali environments. Lowest corrosive aggressiveness has been found for CsOH followed by KOH. Concluding from the corrosion behaviour in the different alkali environments and taking into account the tendency to promote accelerate corrosion, CsOH and KOH are possible alternate alkalis for PWR application. (author). 17 refs, 15 figs, 5 tabs

  19. Influence of alkali metal hydroxides on corrosion of Zr-based alloys

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Y H [Korea Atomic Energy Research Inst., Dae Jun (Korea, Republic of); Ruhmann, H; Garzarolli, F [Siemens-KWU, Power Generation Group, Erlangen (Germany)

    1997-02-01

    In this study the influence of group-1 alkali hydroxides on different zirconium based alloys has been evaluated. The experiments have been carried out in small stainless steel autoclaves at 350 deg. C in pressurized 17 MPa water, with in low (0.32 mmol), medium (4.3 mmol) and high (31.5 mmol) equimolar concentrations of Li-, Na-, K-, Rb- and Cs-Hydroxides. Two types of alloys have been investigated: Zr-Sn-(Transition metal) and Zr-Sn-Nb-(Transition metal). The corrosion behaviour was evaluated from weight gain measurements. From the experiments the cation could be identified as the responsible species for zirconium alloy corrosion in alkalized water. The radius of the cation governs the corrosion behaviour in the pre accelerated region of zircaloy corrosion. Incorporating of alkali cations into the zirconium oxide lattice is probably the mechanism which allows the corrosion enhancement for Li and Na and the significantly lower effect for the other bases. Nb containing alloys show lower corrosion resistance than alloys from the Zr-Sn-TRM system in all alkali solutions. Both types of alloys corrode significantly more in LiOH and NaOH than in the other alkali environments. Lowest corrosive aggressiveness has been found for CsOH followed by KOH. Concluding from the corrosion behaviour in the different alkali environments and taking into account the tendency to promote accelerate corrosion, CsOH and KOH are possible alternate alkalis for PWR application. (author). 17 refs, 15 figs, 5 tabs.

  20. Study on manufacturing technology of fuel guide tube using HANA alloys

    International Nuclear Information System (INIS)

    Kim, Hyungil; Jung, Yangil; Park, Dongjun; Park, Jeongyong; Kim, Ilhyun; Choi, Byungkwon; Jeong, Yonghwan; Park, Sangyoon

    2013-04-01

    This research was focused on the study for the manufacturing technology of HANA alloys to crease the corrosion resistance of 30% as well as the to improve the strength of 10% when compared to the commercial zirconium alloys. The new manufacturing concept having higher corrosion resistance and strength than commercial alloy performance can be obtained in this research. This result was transferred to the KNF and, that will be commercialized. This research result can be summarized like this; Ο Parameter study to increase formability of HANA alloy tube - Study on alloy element and heat-treatment effect - Study on texture development mechanism - Study on final annealing effect Ο Out-of-pile performance evaluation of HANA alloy tube - Corrosion performance evaluation of HANA alloy manufactured at KNF - Mechanical performance evaluation of HANA alloy manufactured at KNF - Recrystallization behavior evaluation of HANA alloy manufactured at KNF - Texture characterization of HANA alloy manufactured at KNF - Microstructure characterization of HANA alloy manufactured at KNF Ο Manufacturing guideline setup to increase formability of HANA alloy tube - Manufacturing guideline setup to decrease surface defect - Manufacturing guideline setup to increase strength and corrosion resistance - Manufacturing guideline setup to control texture

  1. A comparative study on the fretting wear properties of advanced zirconium fuel cladding materials

    International Nuclear Information System (INIS)

    Lee, Young Ho; Kim, Hyung Kyu; Park, Jeong Yong; Kim, Jun Hwan

    2005-06-01

    Fretting wear tests were carried out in room and high temperature water in order to evaluate the wear properties of new zirconium nuclear fuel claddings (K2∼K6) and the commercial claddings (M5, zirlo and zircaloy-4). The objective is to compare the wear resistance of K2∼K6 claddings with that of the commercial ones at the same test condition. After the wear tests, the average wear volume and the maximum wear depth were evaluated and compared at each test condition. As a result, it is difficult to select the most wear-resistant cladding between the K2∼K6 claddings and the commercial ones. This is because the average wear volume and maximum depth of each cladding included between the scattering range of measured results. However, wear resistance of the tested claddings based on the average wear volume and maximum wear depth could be summarized as follows: K5 > zircaloy-4 > (K2,K3) > (K4,M5) > K6 > zirlo at room temperature, zircaloy-4 > K5 > (K3,K4,zirlo) > (K2,K6) > M5 at high temperature and pressure. Therefore, it is concluded that K5 cladding among the tested new zirconium alloys has relatively higher wear-resistance in room and high temperature condition. In order to examine the wear mechanism, it is necessary to systematically study with the consideration of the alloying element effect and test environment. In this report, the wear test procedure and the wear evaluation method are described in detail

  2. Inhibitors for the corrosion of reactive metals: titanium and zirconium and their alloys in acid media

    International Nuclear Information System (INIS)

    Petit, J.A.; Chatainier, G.; Dabosi, F.

    1981-01-01

    The search for effective corrosion inhibitors for titanium and zirconium in acid media is growing because of the considerable increase in the use of these materials in chemical process equipment. It still remains limited, as appears from this review, because of the exceptionally high corrosion resistance of the metals. Titanium has received the greater attention. Its corrosion rate can be lowered by introduction in the medium of multivalent ions, inorganic and organic oxidants. Care should be taken to hold the concentration at a level exceeding some critical value, otherwise the corrosion rate increases. Complexing organic agents do not show such hazardous behaviour. The very rapid corrosion of titanium and zirconium in fluoride media may be lessened by complexing the fluoride ions. Though rarely encountered, localized corrosion may be avoided by using inhibitors. In some cases good corrosion inhibitors for titanium are dissolution accelerators for zirconium. (author)

  3. Synthesis of zirconium by zirconium tetrachloride reduction by magnesio-thermia. Experimental study and modelling

    International Nuclear Information System (INIS)

    Basin, N.

    2001-01-01

    This work deals with the synthesis of zirconium. The ore is carbo-chlorinated to obtain the tetrachloride which is then purified by selective condensation and extractive distillation. Zirconium tetrachloride is then reduced by magnesium and the pseudo-alloy is obtained according to the global following reaction (Kroll process): ZrCl 4 + 2 Mg = 2 MgCl 2 . By thermodynamics, it has been shown that the volatilization of magnesium chloride and the formation of zirconium sub-chlorides are minimized when the combined effects of temperature and of dilution with argon are limited. With these conditions, the products, essentially zirconium and magnesium chloride, are obtained in equivalence ratio in the magnesio-thermia reaction. The global kinetics of the reduction process has been studied by a thermal gravimetric method. A thermo-balance device has been developed specially for this kinetics study. It runs under a controlled atmosphere and is coupled to a vapor tetrachloride feed unit. The transformation is modelled supposing that the zirconium and magnesium chloride formation result: 1)of the evaporation of magnesium from its liquid phase 2)of the transfer of magnesium and zirconium tetrachloride vapors towards the front of the reaction located in the gaseous phase 3)of the chemical reaction. In the studied conditions, the diffusion is supposed to be the limiting process. The influence of the following parameters: geometry of the reactive zone, temperature, scanning rate of the argon-zirconium tetrachloride mixture, composition of the argon-zirconium tetrachloride mixture has been experimentally studied and confronted with success to the model. (O.M.)

  4. Experimental approach and micro-mechanical modeling of the mechanical behavior of irradiated zirconium alloys; Approche experimentale et modelisation micromecanique du comportement des alliages de zirconium irradies

    Energy Technology Data Exchange (ETDEWEB)

    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)

  5. 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

  6. Absorption of dissolved hydrogen from lithiated water during accelerated corrosion of zirconium-2.5 wt% niobium alloy

    International Nuclear Information System (INIS)

    Manolescu, A.V.; Mayer, P.; Rasile, E.M.; Mummenhoff, J.W.

    1982-01-01

    A series of laboratory experiments was carried out to determine the extent of dissolved hydrogen absorption from lithiated water by zirconium-2.5 wt% niobium alloy during corrosion. The material was exposed at 340 0 C to 1 M LiOH aqueous solution containing 0 to approximately 70 cm 3 /L of dissolved hydrogen. Results indicate that dissolved hydrogen has no effect on the corrosion rate or on the amount of hydrogen absorbed by the material

  7. Contribution of in situ acoustic emission analysis coupled with thermogravimetry to study zirconium alloy oxidation

    International Nuclear Information System (INIS)

    Al Haj, O.; Peres, V.; Serris, E.; Cournil, M.; Grosjean, F.; Kittel, J.; Ropital, F.

    2015-01-01

    Zirconium alloy (zircaloy-4) corrosion behavior under oxidizing atmosphere at high temperature was studied using thermogravimetric experiment associated with acoustic emission analysis. Under a mixture of oxygen and air in helium, an acceleration of the corrosion is observed due to the detrimental effect of nitrogen which produces zirconium nitride. The kinetic rate increases significantly after a kinetic transition (breakaway). This acceleration is accompanied by an acoustic emission (AE) activity. Most of the acoustic emission bursts were recorded after the kinetic transition or during the cooling of the sample. Acoustic emission signals analysis allows us to distinguish different populations of cracks in the ZrO 2 layer. These cracks have also been observed by SEM on post mortem cross section of oxidized samples and by in-situ microscopy observations on the top surface of the sample during oxidation. The numerous small convoluted thin cracks observed deeper in the zirconia scale are not detected by the AE technique. From these studies we can conclude that mechanisms as irreversible mechanisms, as cracks initiation and propagation, generate AE signals

  8. Hexagonal close packed to face centered cubic polymorphic transformation in nanocrystalline titanium-zirconium system by mechanical alloying

    International Nuclear Information System (INIS)

    Bera, S.; Manna, I.

    2006-01-01

    The present study reports a reversible hexagonal close packed (hcp) to face centered cubic (fcc) polymorphic phase transformation in four different nanocrystalline titanium-zirconium binary alloys in the course of mechanical alloying in a planetary ball mill. This transformation is monitored at appropriate stages by X-ray diffraction and high-resolution transmission electron microscopy. Lattice parameter of the nanocrystalline fcc phase is a function of the alloy composition. For a given alloy, the lattice parameter and hence volume per atom increase with increase in milling time under comparable conditions. On the other hand, crystallite size, measured from X-ray peak broadening, significantly decreases with the progress of milling. It is suggested that structural instability due to plastic strain, increasing lattice expansion, and negative (from core to boundary) hydrostatic pressure is responsible for this hcp → fcc polymorphic transformation. The said transformation seems reversible as isothermal annealing at 1000 deg. C for 1 h or melting the powder mass leads to partial or complete transformation of the milled product from single phase fcc to hcp

  9. Defect kinetics and resistance to amorphization in zirconium carbide

    International Nuclear Information System (INIS)

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

    2015-01-01

    To better understand the radiation response of zirconium carbide (ZrC), and in particular its excellent resistance to amorphization, we have used density functional theory methods to study the kinetics of point defects in ZrC. The migration barriers and recombination barriers of the simple point defects are calculated using the ab initio molecular dynamics simulation and the nudged elastic band method. These barriers are used to estimate C and Zr interstitial and vacancy diffusion and Frenkel pair recombination rates. A significant barrier for C Frenkel pair recombination is found but it is shown that a large concentration of C vacancies reduces this barrier dramatically, allowing facile healing of radiation damage. The mechanisms underlying high resistance to amorphization of ZrC were analyzed from the perspectives of structural, thermodynamic, chemical and kinetic properties. This study provides insights into the amorphization resistance of ZrC as well as a foundation for understanding general radiation damage in this material

  10. Improving Accident Tolerance of Nuclear Fuel with Coated Mo-alloy Cladding

    Directory of Open Access Journals (Sweden)

    Bo Cheng

    2016-02-01

    Full Text Available In severe loss of coolant accidents (LOCA, similar to those experienced at Fukushima Daiichi and Three Mile Island Unit 1, the zirconium alloy fuel cladding materials are rapidly heated due to nuclear decay heating and rapid exothermic oxidation of zirconium with steam. This heating causes the cladding to rapidly react with steam, lose strength, burst or collapse, and generate large quantities of hydrogen gas. Although maintaining core cooling remains the highest priority in accident management, an accident tolerant fuel (ATF design may extend coping and recovery time for operators to restore emergency power, and cooling, and achieve safe shutdown. An ATF is required to possess high resistance to steam oxidation to reduce hydrogen generation and sufficient mechanical strength to maintain fuel rod integrity and core coolability. The initiative undertaken by Electric Power Research Institute (EPRI is to demonstrate the feasibility of developing an ATF cladding with capability to maintain its integrity in 1,200–1,500°C steam for at least 24 hours. This ATF cladding utilizes thin-walled Mo-alloys coated with oxidation-resistant surface layers. The basic design consists of a thin-walled Mo alloy structural tube with a metallurgically bonded, oxidation-resistant outer layer. Two options are being investigated: a commercially available iron, chromium, and aluminum alloy with excellent high temperature oxidation resistance, and a Zr alloy with demonstrated corrosion resistance. As these composite claddings will incorporate either no Zr, or thin Zr outer layers, hydrogen generation under severe LOCA conditions will be greatly reduced. Key technical challenges and uncertainties specific to Mo alloy fuel cladding include: economic core design, industrial scale fabricability, radiation embrittlement, and corrosion and oxidation resistance during normal operation, transients, and severe accidents. Progress in each aspect has been made and key results are

  11. Oxidation resistant chromium coating on Zircaloy-4 for accident tolerant fuel cladding

    International Nuclear Information System (INIS)

    Park, Jung-Hwan; Kim, Eui-Jung; Jung, Yang-Il; Park, Dong-Jun; Kim, Hyun-Gil; Park, Jeong-Yong; Koo, Yang-Hyun

    2015-01-01

    The attributes of such a fuel are approved reaction kinetics with steam, a slower hydrogen generation rate, and good cladding thermo-mechanical properties. Many researchers have tried to modify zirconium alloys to improve their oxidation resistance in the early stages of the ATF development. Corrosion resistant coating on cladding is one of the candidate technologies to improve the oxidation resistance of zirconium cladding. By applying coating technology to zirconium cladding, it is easy to obtain corrosion resistance without a change in the base materials. Among the surface coating methods, arc ion plating (AIP) is a coating technology to improve the adhesion owing to good throwing power, and a dense deposit (Fig. 1). Owing to these advantages, AIP has been widely used to efficiently form protective coatings on cutting tools, dies, bearings, etc. In this study, The AIP technique for the protection of zirconium claddings from the oxidation in a high-temperature steam environment was studied. The homogeneous Cr film with a high adhesive ability to the cladding was deposited by AIP and acted as a protection layer to enhance the corrosion resistance of the zirconium cladding. It was concluded that the AIP technology is effective for coating a protective layer on claddings

  12. Oxidation resistant chromium coating on Zircaloy-4 for accident tolerant fuel cladding

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jung-Hwan; Kim, Eui-Jung; Jung, Yang-Il; Park, Dong-Jun; Kim, Hyun-Gil; Park, Jeong-Yong; Koo, Yang-Hyun [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

    The attributes of such a fuel are approved reaction kinetics with steam, a slower hydrogen generation rate, and good cladding thermo-mechanical properties. Many researchers have tried to modify zirconium alloys to improve their oxidation resistance in the early stages of the ATF development. Corrosion resistant coating on cladding is one of the candidate technologies to improve the oxidation resistance of zirconium cladding. By applying coating technology to zirconium cladding, it is easy to obtain corrosion resistance without a change in the base materials. Among the surface coating methods, arc ion plating (AIP) is a coating technology to improve the adhesion owing to good throwing power, and a dense deposit (Fig. 1). Owing to these advantages, AIP has been widely used to efficiently form protective coatings on cutting tools, dies, bearings, etc. In this study, The AIP technique for the protection of zirconium claddings from the oxidation in a high-temperature steam environment was studied. The homogeneous Cr film with a high adhesive ability to the cladding was deposited by AIP and acted as a protection layer to enhance the corrosion resistance of the zirconium cladding. It was concluded that the AIP technology is effective for coating a protective layer on claddings.

  13. Casting characteristics of Al-Mg alloy 535 cast in permanent moulds

    International Nuclear Information System (INIS)

    Fasoyinu, F.A.; Thomson, J.; Cousineau, D.; Castles, T.; Sahoo, M.

    2002-01-01

    Aluminum alloy 535 could be used for automotive and marine applications because of its good corrosion resistance against mild alkaline and salt spray exposure. The majority of components from this alloy are usually produced by sand casting because it is prone to hot shortness and has poor fluidity when poured in permanent moulds. In an attempt to improve its castability in permanent moulds, casting characteristics such as casting fluidity and hot tear resistance have been studied. In addition, the effectiveness of titanium, boron, scandium, zirconium and a combination of selected elements from this group as grain refiners were evaluated. It s shown that alloy 535 exhibits good casting fluidity when poured with adequate metal superheat and that there is significant improvement in hot tear resistance following grain refinement. (author)

  14. Identification and characterization of a new Zirconium hydride

    International Nuclear Information System (INIS)

    Zhao, Z.

    2007-01-01

    In order to control the integrity of the fuel clad, alloy of zirconium, it is necessary to predict the behavior of zirconium hydrides in the environment (temperature, stress...), at a microscopic scale. A characterization study by TEM of hydrides has been realized. It shows little hydrides about 500 nm, in hydride Zircaloy 4. Then a more detailed study identified a new hydride phase presented in this paper. (A.L.B.)

  15. Effect of zirconium addition on the ductility and toughness of cast zinc-aluminum alloy5, zamak5, grain refined by titanium plus boron

    International Nuclear Information System (INIS)

    Adnan, I.O.

    2007-01-01

    Zinc-aluminum casting alloys are frequently employed in design. They are inexpensive and have mechanical properties in many respects superior to aluminum and copper alloys. Common applications of zinc-aluminum alloys are in the automobile industry for manufacturing carburetors bodies, fuel pump bodies, driving wheels and door handles. They are mainly used for die casting due to their low melting points which ranges from 375 to 487 degree C, good fluidity, pollution free melting in addition to their high corrosion resistance. Against these advantages there exists the deficiency as these alloys solidify in a coarse dentititic structure which tends to deteriorate the mechanical properties and impact strength. It was found that addition of some rare earth materials e.g. titanium or titanium plus boron results in modifying its structure into a petal-like or nodular type. The available literature reveals that most of the published work is directed towards the metallurgical aspects and little or no work is published on the effect of those elements on its mechanical strength, ductility, toughness and impact strength. In this paper, the effect of addition of Zirconium on the microstructure, mechanical behavior, hardness, ductility and impact strength of zinc-aluminum alloy5, Zamak5, is investigated. It was found that addition of Ti+B or Zr or Ti+B+Zr resulted in modifying the coarse dentritic structure of the Zamak5 alloy into a fine nodular one. Further more, addition of any of these elements alone or together resulted in enhancement of the mechanical strength, hardness, ductility, toughness and impact strength of this alloy, for example an increase of 11% in hardness was achieved in case of Zr addition and 100% increase of ductility and 12.5% increase in impact strength were achieved in case of Ti+B addition. (author)

  16. Oxidation kinetics of some zirconium alloys in flowing carbon dioxide at high temperatures

    International Nuclear Information System (INIS)

    Kohli, R.

    1980-01-01

    The oxidation kinetics of three zirconium alloys (Zr-2.2 wt% Hf, Zr-2.5 wt% Nb, and Zr-3 wt% Nb-1 wt% Sn) have been measured in flowing carbon dioxide in the temperature range from 873 to 1173 K to 120 ks (2000 min). At all oxidation temperatures, Zr-2.5 Nb and Zr-3 Nb-1 Sn showed a transition to rapid linear kinetics after initial parabolic oxidation. The Zr-2.2 Hf showed this transition at temperatures in the range from 973 to 1173 K; at 873 K, no transition was observed within the oxidation times reported. The Zr-2.2 Hf showed the smallest weight gains, followed in order by Zr-2.5 Nb and Zr-3 Nb-1 Sn. Increased oxidation rates and shorter times-to-rate-transition of Zr-2.2 Nb and Zr-1 Sn as compared with Zr-2.2 Hf can be attributed to the presence of niobium, tin, and hafnium in the alloys. This is considered in terms of the Nomura-Akutsu model, according to which hafnium should delay the rate transition, while niobium and tin lead to shorter times-to-rate-transition. The scale on Zr-2.2 Hf was identified as monoclinic zirconia, while the tetragonal phase, 6ZrO 2 .Nb 2 O 5 , was contained in the monoclinic zirconia scales on both other alloys

  17. Corrosion performance of new Zircaloy-2-based alloys

    International Nuclear Information System (INIS)

    Rudling, P.; Mikes-Lindbaeck, M.; Lethinen, B.; Andren, H.O.; Stiller, K.

    1994-01-01

    A material development project was initiated to develop a new zirconium alloy, outside the ASTM specifications for Zircaloy-2 and Zircaloy-4, with optimized hydriding and corrosion properties for both boiling water reactors and pressurized water reactors. A number of different alloys were manufactured. These alloys were long-term corrosion tested in autoclaves at 400 C in steam. Also, a 520 C/24 h steam test was carried out. The zirconium metal microstructure and the chemistry of precipitates were characterized by analytical electron microscopy. The metal matrix chemistry was determined by atom probe analysis. The paper describes the correlations between corrosion material performance and zirconium alloy microstructure

  18. METHOD OF IMPROVING CORROSION RESISTANCE OF ZIRCONIUM

    Science.gov (United States)

    Shannon, D.W.

    1961-03-28

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

  19. Argon-arc welding of heat resisting aluminium alloys

    International Nuclear Information System (INIS)

    Ryazantsev, V.I.; Fedoseev, V.A.

    1997-01-01

    Welding of aluminium heat resisting alloys of the Al-Cu-Mg system is studied. The hot-shortness of heat-resistant alloys M40, 1150 and 1151 are at the level of aluminium alloys 1201 and by 2-3 times lower as compared to the aluminium alloy AMg6. The M40, 1150 and 1151 alloys have unquestionable advantages against other know aluminium alloys only at temperatures of welded structures operation, beginning with 150-2000 deg C and especially at 250 deg C

  20. Passivation of mechanically polished, chemically etched and anodized zirconium in various aqueous solutions: Impedance measurements

    International Nuclear Information System (INIS)

    Abo-Elenien, G.M.; Abdel-Salam, O.E.

    1987-01-01

    Zirconium and its alloys are finding increasing applications especially in water-cooled nuclear reactors. Because of the fact that zirconium is electronegative (E 0 = -1.529V) its corrosion resistance in aqueous solutions is largely determined by the existence of a thin oxide film on its surface. The structure and properties of this film depend in the first place on the method of surface pre-treatment. This paper presents an experimental study of the nature of the oxide film on mechanically polished, chemically etched and anodized zirconium. Ac impedance measurements carried out in various acidic, neutral and alkaline solutions show that the film thickness depends on the method of surface pre-treatment and the type of electrolyte solution. The variation of the potential and impedance during anodization of zirconium at low current density indicates that the initial stages of polarization consist of oxide build-up at a rate dependent on the nature of the electrode surface and the electrolyte. Oxygen evolution commences at a stage where oxide thickening starts to decline. The effect of frequency on the measured impedance indicates that the surface reactivity, and hence the corrosion rate, decreases in the following order: mechanically polished > chemically etched > anodized

  1. KTA 625 alloy tube with excellent corrosion resistance and heat resistance

    International Nuclear Information System (INIS)

    Fujiwara, Kazuo; Kadonaga, Toshiki; Kikuma, Seiji.

    1982-01-01

    The problems when seamless tubes are produced by using nickel base 625 alloy (61Ni-22Cr-9Mo-Cb) which is known as a corrosion resistant and heat resistant alloyF were examined, and the confirmation experiment was carried out on its corrosion resistance and heat resistance. Various difficulties have been experienced in the tube making owing to the characteristics due to the chemical composition, but they were able to be solved by the repeated experiments. As for the characteristics of the product, the corrosion resistance was excellent particularly in the environment containing high temperature, high concentration chloride, and also the heat resistance was excellent in the wide temperature range from normal temperature to 1000 deg C. From these facts, the wide fields of application are expected for these alloy tubes, including the evaporation and concentration equipment for radioactive wastes in atomic energy field. Expecting the increase of demand hereafter, Kobe Steel Ltd. examined the problems when seamless tubes are produced from the 625 alloy by Ugine Sejournet process. The aptitude for tube production such as the chemical composition, production process and the product characteristics, the corrosion resistance against chloride, hydrogen sulfide, polythionic and other acids,F the high temperature strength and oxidation resistance are reported. (Kako, I.)

  2. Grain refinement of permanent mold cast copper base alloys. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Sadayappan, M.; Thomson, J. P.; Elboujdaini, M.; Gu, G. Ping; Sahoo, M.

    2004-04-29

    Grain refinement behavior of copper alloys cast in permanent molds was investigated. This is one of the least studied subjects in copper alloy castings. Grain refinement is not widely practiced for leaded copper alloys cast in sand molds. Aluminum bronzes and high strength yellow brasses, cast in sand and permanent molds, were usually fine grained due to the presence of more than 2% iron. Grain refinement of the most common permanent mold casting alloys, leaded yellow brass and its lead-free replacement EnviroBrass III, is not universally accepted due to the perceived problem of hard spots in finished castings and for the same reason these alloys contain very low amounts of iron. The yellow brasses and Cu-Si alloys are gaining popularity in North America due to their low lead content and amenability for permanent mold casting. These alloys are prone to hot tearing in permanent mold casting. Grain refinement is one of the solutions for reducing this problem. However, to use this technique it is necessary to understand the mechanism of grain refinement and other issues involved in the process. The following issues were studied during this three year project funded by the US Department of Energy and the copper casting industry: (1) Effect of alloying additions on the grain size of Cu-Zn alloys and their interaction with grain refiners; (2) Effect of two grain refining elements, boron and zirconium, on the grain size of four copper alloys, yellow brass, EnviroBrass II, silicon brass and silicon bronze and the duration of their effect (fading); (3) Prediction of grain refinement using cooling curve analysis and use of this method as an on-line quality control tool; (4) Hard spot formation in yellow brass and EnviroBrass due to grain refinement; (5) Corrosion resistance of the grain refined alloys; (6) Transfer the technology to permanent mold casting foundries; It was found that alloying elements such as tin and zinc do not change the grain size of Cu-Zn alloys

  3. A computer model for hydride blister growth in zirconium alloys

    International Nuclear Information System (INIS)

    White, A.J.; Sawatzky, A.; Woo, C.H.

    1985-06-01

    The failure of a Zircaloy-2 pressure tube in the Pickering unit 2 reactor started at a series of zirconium hydride blisters on the outside of the pressure tube. These blisters resulted from the thermal diffusion of hydrogen to the cooler regions of the pressure tube. In this report the physics of thermal diffusion of hydrogen in zirconium is reviewed and a computer model for blister growth in two-dimensional Cartesian geometry is described. The model is used to show that the blister-growth rate in a two-phase zirconium/zirconium-hydride region does not depend on the initial hydrogen concentration nor on the hydrogen pick-up rate, and that for a fixed far-field temperature there is an optimum pressure-type/calandria-tube contact temperature for growing blisters. The model described here can also be used to study large-scale effects, such as hydrogen-depletion zones around hydride blisters

  4. Roles of texture of Zr alloys in ZrO{sub 2} film formation and δ-hydride orientation near ZrO{sub 2}/Zr interface

    Energy Technology Data Exchange (ETDEWEB)

    Qin, W.; Szpunar, J.A., E-mail: weq565@mail.usask.ca, E-mail: jerzy.szpunar@usask.ca [Univ. of Saskatchewan, Dept. of Mechanical Engineering, Saskatoon, SK (Canada); Kozinski, J., E-mail: janusz.kozinski@lassonde.yorku.ca [York Univ., Faculty of Science and Engineering, Toronto, ON (Canada)

    2014-07-01

    Oxidation and hydrogen embrittlement are related to formation of cracks and failure of Zr alloys used in nuclear reactor applications. An in-depth understanding of the formation of ZrO{sub 2} film and the hydride precipitation and orientation is important for improving the corrosion resistance of zirconium alloys. In this work a theoretical model is developed to analyze the microstructure of ZrO{sub 2} film formed on Zr alloys and the effect of stress that results from ZrO{sub 2} formation on hydride reorientation in the region near oxide/metal interface. Our work shows that the macroscopic stress produced due to Pilling-Bedworth ratio for ZrO{sub 2}/Zr could lead to the hydride re-orientation in the region near ZrO{sub 2}/Zr interface. Whether or not this effect can occur is dependent on the texture of the zirconium alloys. Control of texture of zirconium alloys can affect the microstructure of ZrO{sub 2} film and can be responsible for change of hydride orientation. (author)

  5. Determination of zirconium by fluoride ion selective electrode

    International Nuclear Information System (INIS)

    Mahanty, B.N.; Sonar, V.R.; Gaikwad, R.; Raul, S.; Das, D.K.; Prakash, A.; Afzal, Md.; Panakkal, J.P.

    2010-01-01

    Full text: Zirconium is used in a wide range of applications including nuclear clad, catalytic converters, surgical appliances, metallurgical furnaces, superconductors, ceramics, lamp filaments, anti corrosive alloys and photographical purposes. Irradiation testing of U-Zr and U-Pu-Zr fuel pins has also demonstrated their feasibility as fuel in liquid metal reactors. Different methods that are employed for the determination of zirconium are spectrophotometry, potentiometry, neutron activation analysis and mass spectrometry. Ion-selective electrode (ISE), selective to zirconium ion has been studied for the direct potentiometric measurements of zirconium ions in various samples. In the present work, an indirect method has been employed for the determination of zirconium in zirconium nitrate sample using fluoride ion selective electrode. This method is based on the addition of known excess amount of fluoride ion to react with the zirconium ion to produce zirconium tetra fluoride at about pH 2-3, followed by determination of residual fluoride ion selective electrode. The residual fluoride ion concentrations were determined from the electrode potential data using calibration plot. Subsequently, zirconium ion concentrations were determined from the concentration of consumed fluoride ions. A precision of about 2% (RSD) with the mean recovery of more than 94% has been achieved for the determination of zirconium at the concentration of 4.40 X 10 -3 moles lit -1

  6. Study on direct dissolution of U-10Zr alloy and distribution of uranium and zirconium in liquid cadmium

    International Nuclear Information System (INIS)

    Ye Yuxing; Gao Yuan

    1997-09-01

    The effect of dissolution time, temperature, total surface area of U-10Zr alloy pellets and stirring on the dissolution and dissolution rate of uranium in liquid cadmium were studied. Cadmium containing U and Zr dissolved from U-10Zr alloy at 475 degree C and 500 degree C respectively was analyzed with electron microanalyzer. The experimental results show that at 400 degree and 500 degree C with the stirring rate of some 150 r/min, the solubilities of uranium in liquid cadmium are 0.4% and 2.2%, respectively. At the first 30 min, the dissolution rates of U-10Zr alloy pellets are 0.05 g/(cm 2 ·h) and 0.32 g/(cm 2 ·h), respectively. The suitable dissolution conditions for U-10Zr alloy pellets in liquid cadmium (the ratio of the mass of liquid cadmium to that of the pellets ≅7) are: temperature, about 480 degree C; stirring rate, about 150 r/min; dissolution time, 4 h. The distribution of uranium and zirconium in cadmium is homogeneous

  7. Investigation of anodic oxide coatings on zirconium after heat treatment

    International Nuclear Information System (INIS)

    Sowa, Maciej; Dercz, Grzegorz; Suchanek, Katarzyna; Simka, Wojciech

    2015-01-01

    Highlights: • Oxide layers prepared via PEO of zirconium were subjected to heat treatment. • Surface characteristics were determined for the obtained oxide coatings. • Heat treatment led to the partial destruction of the anodic oxide layer. • Pitting corrosion resistance of zirconium was improved after the modification. - Abstract: Herein, results of heat treatment of zirconium anodised under plasma electrolytic oxidation (PEO) conditions at 500–800 °C are presented. The obtained oxide films were investigated by means of SEM, XRD and Raman spectroscopy. The corrosion resistance of the zirconium specimens was evaluated in Ringer's solution. A bilayer oxide coatings generated in the course of PEO of zirconium were not observed after the heat treatment. The resulting oxide layers contained a new sublayer located at the metal/oxide interface is suggested to originate from the thermal oxidation of zirconium. The corrosion resistance of the anodised metal was improved after the heat treatment

  8. Amine extraction of lead(II) and zirconium(IV) with succinate media

    International Nuclear Information System (INIS)

    Mahamuni, S.V.; Mane, C.P.; Sargar, B.M.; Rajmane, M.M.; Anuse, M.A.

    2004-01-01

    Lead is an important constituent of various alloys, which are in increasing demand in manufacture of batteries and nuclear shielding while the use of zirconium in nuclear power plants as entirely cladding uranium fuel is most important. This study was carried out to optimize the extraction conditions for Pb(II) and zirconium(IV)

  9. Delayed hydride cracking of zirconium alloy fuel cladding

    International Nuclear Information System (INIS)

    2010-10-01

    This report describes the work performed in a coordinated research project on Hydrogen and Hydride Degradation of the Mechanical and Physical Properties of Zirconium Alloys. It is the second in the series. In 2005-2009 that work was extended within a new CRP called Delayed Hydride Cracking in Zirconium Alloy Fuel Cladding. The project consisted of adding hydrogen to samples of Zircaloy-4 claddings representing light water reactors (LWRs), CANDU and Atucha, and measuring the rates of delayed hydride cracking (DHC) under specified conditions. The project was overseen by a supervisory group of experts in the field who provided advice and assistance to participants as required. All of the research work undertaken as part of the CRP is described in this report, which includes details of the experimental procedures that led to a consistent set of data for LWR cladding. The participants and many of their co-workers in the laboratories involved in the CRP contributed results and material used in this report, which compiles the results, their analysis, discussions of their interpretation and conclusions and recommendations for future work. The research was coordinated by an advisor and by representatives in three laboratories in industrialized Member States. Besides the basic goal to transfer the technology of the testing technique from an experienced laboratory to those unfamiliar with the methods, the CRP was set up to harmonize the experimental procedures to produce consistent sets of data, both within a single laboratory and between different laboratories. From the first part of this project it was demonstrated that by following a standard set of experimental protocols, consistent results could be obtained. Thus, experimental vagaries were minimized by careful attention to detail of microstructure, temperature history and stress state in the samples. The underlying idea for the test programme was set out at the end of the first part of the project on pressure tubes. The

  10. Recent irradiation tests of uranium-plutonium-zirconium metal fuel elements

    International Nuclear Information System (INIS)

    Pahl, R.G.; Lahm, C.E.; Villarreal, R.; Hofman, G.L.; Beck, W.N.

    1986-09-01

    Uranium-Plutonium-Zirconium metal fuel irradiation tests to support the ANL Integral Fast Reactor concept are discussed. Satisfactory performance has been demonstrated to 2.9 at.% peak burnup in three alloys having 0, 8, and 19 wt % plutonium. Fuel swelling measurements at low burnup in alloys to 26 wt % plutonium show that fuel deformation is primarily radial in direction. Increasing the plutonium content in the fuel diminishes the rate of fuel-cladding gap closure and axial fuel column growth. Chemical redistribution occurs by 2.1 at.% peak burnup and generally involves the inward migration of zirconium and outward migration of uranium. Fission gas release to the plenum ranges from 46% to 56% in the alloys irradiated to 2.9 at.% peak burnup. No evidence of deleterious fuel-cladding chemical or mechanical interaction was observed

  11. Structural studies of calcium phosphate doped with titanium and zirconium obtained by high-energy mechanical alloying

    Energy Technology Data Exchange (ETDEWEB)

    Silva, C C; Sombra, A S B [Telecommunications and Materials Science and Engineering Laboratory (LOCEM), Physics Department, Federal University of Ceara, Campus do Pii, Postal Code 6030, 60455-760, Fortaleza-Ceara (Brazil)], E-mail: sombra@fisica.ufc.br

    2009-12-15

    In this paper, we present a new variation of the solid-state procedure on the synthesis of bioceramics with titanium (CapTi) and zirconium (CapZr), considering that zirconium (ZrO{sub 2}) and titanium oxide (TiO{sub 2}) are strengthening agents, due to their superb force and fracture toughness. The high efficiency of the calcination process opens a new way of producing commercial amounts of nanocrystalline bioceramics. In this work, a new variation of the solid-state procedure method was used to produce nanocrystalline powders of titanium and zirconium, using two different experimental chemical routes: CapTi: Ca(H{sub 2}PO{sub 4}){sub 2}+TiO{sub 2} and CapZr: Ca(H{sub 2}PO{sub 4}){sub 2}+ZrO{sub 2}. The powders were submitted to calcination processes (CapTic and CapZrc) at 800, 900 and 1000 deg. C. The calcium titanium phosphate phase, CaTi{sub 4}P{sub 6}O{sub 24}, was obtained in the CapTic reaction and the calcium zirconium phosphate, CaZr{sub 4}P{sub 6}O{sub 24}, was obtained in the CapZrc reaction. The obtained ceramics were characterized by x-ray powder diffraction (XRD), infrared (IR) spectroscopy, Raman scattering spectroscopy (RSS) and scanning electron microscopy (SEM) analysis. This method was compared with the milling process (CapTim and CapZrm), where in the last process the melting is not necessary and the powder obtained is nanocrystalline. The calcium titanium phosphate phase, CaTi{sub 4}P{sub 6}O{sub 24}, was obtained in the reaction CapTim, but in CapZrm the formation of any calcium phosphate phase even after 15 h of dry mechanical alloying was not observed.

  12. Susceptibility of cold-worked zirconium-2.5 wt% niobium alloy to delayed hydrogen cracking

    International Nuclear Information System (INIS)

    Coleman, C.E.

    1976-01-01

    Notched tensile specimens of cold-worked zirconium-2.5 wt% niobium alloy have been stressed at 350 K and 520 K. At 350 K, above a possible threshold stress of 200 MPa, specimens exhibited delayed failure which was attributed to hydride cracking. Metallography showed that hydrides accumulated at notches and tips of growing cracks. The time to failure appeared to be independent of hydrogen content over the range 7 to 100 ppm hydrogen. Crack growth rates of about 10 -10 m/s deduced from fractography were in the same range as those necessary to fracture pressure tubes. The asymptotic stress intensity for delayed failure, Ksub(1H), appeared to be about 5 MPa√m. With this low value of Ksub(1H) small surface flaws may propagate in pressure tubes which contain large residual stresses. Stress relieving and modified rolling procedures will reduce the residual stresses to such an extent that only flaws 12% of the wall thickness or greater will grow. At 520 K no failures were observed at times a factor of three greater than times to failure at 350 K. Zirconium-2.5 wt% niobium appears to be safe from delayed hydrogen cracking at the reactor operating temperature. (author)

  13. Modelling zirconium hydrides using the special quasirandom structure approach

    KAUST Repository

    Wang, Hao; Chroneos, Alexander I.; Jiang, Chao; Schwingenschlö gl, Udo

    2013-01-01

    The study of the structure and properties of zirconium hydrides is important for understanding the embrittlement of zirconium alloys used as cladding in light water nuclear reactors. Simulation of the defect processes is complicated due to the random distribution of the hydrogen atoms. We propose the use of the special quasirandom structure approach as a computationally efficient way to describe this random distribution. We have generated six special quasirandom structure cells based on face centered cubic and face centered tetragonal unit cells to describe ZrH2-x (x = 0.25-0.5). Using density functional theory calculations we investigate the mechanical properties, stability, and electronic structure of the alloys. © the Owner Societies 2013.

  14. Corrosion behaviour of zirconium alloys in the autoclaves of Embalse nuclear power plant

    International Nuclear Information System (INIS)

    Bordoni, Roberto A.; Olmedo, Ana M.; Villegas, Marina; Miyagusuku, Marcela; Maroto, Alberto J. G.; Sainz, Ricardo A.; Fernandez, Alberto N.; Allemandi, Walter D.

    1999-01-01

    The corrosion behaviour of zirconium alloys coupons attached to the holders of the autoclaves located out of core in the primary circuit of Embalse nuclear power plant is described. The Zr-2.5 Nb coupons of the autoclaves at the higher temperature (305 C degrees) and the Zry-4 coupons of the autoclaves at 265 and 305 C degrees installed in 1988 had a normal corrosion behaviour, after 3500 of full power days. While, the Zr-2.5 Nb coupons, at 265 C degrees, showed the presence of white oxide nuclei and a weight gain indicating an abnormal corrosion behaviour which might be attributed to the material microstructure. Complementary tests, made in the period September 1991-April 1993, showed that the abnormal corrosion behaviour observed for the Canadian coupons installed in 1983 was due to a surface contamination of the Zry-4 coupons and due to the microstructure of the Zr-2.5 Nb coupons. The normal corrosion behaviour for both alloys installed in 1986, showed that the resin ingress to the primary circuit that occurred in 1988, do not affect the performance of these materials. (author)

  15. Lithium uptake and the corrosion of zirconium alloys in aqueous lithium hydroxide solutions

    International Nuclear Information System (INIS)

    Ramasubramanian, N.

    1991-01-01

    This paper reports on corrosion films on zirconium alloys that were analyzed for lithium by Atomic Absorption Spectroscopy (AAS), Secondary Ion Mass Spectrometry (SIMS), and Infrared Reflection Absorption Spectroscopy (IRAS). The oxides grown in reactor in dilute lithium hydroxide solution, specimens cut from Zircaloy, and Zr-2.5Nb alloy pressure tubes removed from CANDU (Canada Deuterium Uranium, Registered Trademark) reactors showed low concentrations of lithium (4 to 50 ppm). The lithium was not leachable in a warm dilute acid. 6 Li undergoes transmutation by the 6 Li(n,t) 4 He reaction. However, SIMS profiles for d 7 Li were identical through the bulk oxide and the isotopic ratio was close to the natural abundance value. The lithium in the oxide, existing as adsorbed lithium on the surface, has been in dynamic equilibrium with lithium in the coolant, and, in spite of many Effective Full Power Years (EFPY) of operation, lithium added to the CANDU coolant at ∼2.5 ppm is not concentrating in the oxides. On the other hand, corrosion films grown in the laboratory in concentrated lithium hydroxide solutions were very porous and contained hundreds of ppm of lithium in the oxide

  16. Corrosion-Resistant High-Entropy Alloys: A Review

    Directory of Open Access Journals (Sweden)

    Yunzhu Shi

    2017-02-01

    Full Text Available Corrosion destroys more than three percent of the world’s gross domestic product. Therefore, the design of highly corrosion-resistant materials is urgently needed. By breaking the classical alloy-design philosophy, high-entropy alloys (HEAs possess unique microstructures, which are solid solutions with random arrangements of multiple elements. The particular locally-disordered chemical environment is expected to lead to unique corrosion-resistant properties. In this review, the studies of the corrosion-resistant HEAs during the last decade are summarized. The corrosion-resistant properties of HEAs in various aqueous environments and the corrosion behavior of HEA coatings are presented. The effects of environments, alloying elements, and processing methods on the corrosion resistance are analyzed in detail. Furthermore, the possible directions of future work regarding the corrosion behavior of HEAs are suggested.

  17. Investigation of in-pile grown corrosion films on zirconium-based alloys

    International Nuclear Information System (INIS)

    Gebhardt, O.; Hermann, A.; Bart, G.; Blank, H.; Ray, I.L.F.

    1996-01-01

    In-pile grown corrosion films on different fuel rod claddings (standard Zircaloy-4, extra low tin Zircaloy (ELS), and Zr2.5Nb) have been studied using a variety of experimental techniques. The aim of the investigations was to find out common features and differences between the corrosion layers grown on zirconium alloys having different composition. Methods applied were scanning and transmission electron microscopy (SEM, TEM), electrochemical impedance spectroscopy (EIS), and electrochemical anodization. The morphological differences have been observed between the specimens that could explain the irradiation enhancement of corrosion of Zircaloy-4. The features of the compact oxide close to the oxide/metal interface have been characterized by electrochemical methods. The relationship between the thickness of this protective oxide and the overall oxide thickness has been investigated by EIS. It was found that this relation is dependent on the location of the oxide along the fuel rod and on the corrosion rate

  18. Effects of surface treatment on the cavitation erosion of high-chrome steel, zirconium, titanium and their alloys

    International Nuclear Information System (INIS)

    Marinin, V.G.

    1994-01-01

    The erosion resistance of some structural materials used for equipment components of the first and second circuits of NPPs is studied under cavitation created by an ultrasonic vibrator. It appears that after various thermomechanical treatments (programmed loading, low-temperature rolling) and coating deposition (titanium, zirconium and titanium nitride), the erosion resistance of the materials under consideration increases and the plasticity value is not notably modified. The titanium coatings deposited onto the steel increase the corrosion-fatigue resistance in a sodium chloride environment, in several cases

  19. Strength of zirconium--titanium martensites and deformation behaviour

    International Nuclear Information System (INIS)

    Banerjee, S.; Vijayakar, S.J.; Krishnan, R.

    1978-01-01

    The deformation behavior of pure zirconium and of zirconium--titanium alloys containing 5, 10, 15 and 20 wt % titanium was studied in two heat treated conditions: furnace cooled and water quenched from the β phase field. By comparing the flow stresses of the furnace cooled α and the water quenched α' (martensite) structures it was possible to isolate the strengthening contributions of the martensitic structure (comprising the contributions due to the small size of the martensite units and to the distributions of defects like dislocations and internal twins) from those arising from the solid solution. The internally twinned plate martensite structure in the Zr--15% Ti and the Zr--20% Ti alloys was responsible for a significant increase in strength, while the strengthening due to the dislocated lath martensite structure in the more dilute alloys was only marginal. Stress relaxation experiments revealed that strengthening associated with the martensite structure was mainly due to an increase in the athermal component of the flow stress. The effectiveness of the lath boundaries and the (10 anti 11) twin boundaries in offering resistance to an approaching deformation front (either slip or twin) was examined. While the lath boundaries were found to be transparent with respect to the propagation of slip dislocations and deformation twins, a majority of plate as well as twin boundaries were effective barriers against their propagation. TEM observations showed an extensive accumulation of geometrically necessary dislocations in the plastically deformed twinned martensites. Enhanced work hardening was related to the geometric slip distances in these structures in accordance with Ashby's one parameter work hardening theory for plastically inhomogeneous materials. The effect of the martensite structure on different components of the flow stress (dependent on or independent of grain size and strain) was discussed

  20. 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)

  1. Microhardness and microplasticity of zirconium nitride

    International Nuclear Information System (INIS)

    Neshpor, V.S.; Eron'yan, M.A.; Petrov, A.N.; Kravchik, A.E.

    1978-01-01

    To experimentally check the concentration dependence of microhardness of 4 group nitrides, microhardness of zirconium nitride compact samples was measured. The samples were obtained either by bulk saturation of zirconium iodide plates or by chemical precipitation from gas. As nitrogen content decreased within the limits of homogeneity of zirconium nitride samples where the concentration of admixed oxygen was low, the microhardness grew from 1500+-100 kg/mm 2 for ZrNsub(1.0) to 27000+-100 kg/mm 2 for ZrNsub(0.78). Microplasticity of zirconium nitride (resistance to fracture) decreased, as the concentration of nitrogen vacancies was growing

  2. Control of microstructure to increase the tolerance of zirconium alloys to hydride cracking

    International Nuclear Information System (INIS)

    Coleman, C.E.; Sagat, S.; Amouzouvi, K.F.

    1987-12-01

    The microstructure of Zr-2.5 Nb has been altered in three ways in attempts to increase the alloy's tolerance to delayed hydride cracking, namely by breaking up the β-phase which reduces diffusivity of hydrogen and decreases crack velocity, by means of a gettering element (yttrium) which reduces susceptibility to cracking although the yttrium alloy has low toughness and poor corrosion resistance, and by reducing the number of basal plane normals in the main stressing direction which improves resistance to crack growth

  3. Multiscale modelling of hydrogen embrittlement in zirconium alloys

    Energy Technology Data Exchange (ETDEWEB)

    Majevadia, Jassel; Wenman, Mark; Balint, Daniel; Sutton, Adrian [Imperial College London (United Kingdom); Nazarov, Roman [MPIE, Dusseldorf (Germany)

    2013-07-01

    Delayed Hydride Cracking (DHC) is a commonly occurring embrittlement phenomenon in zirconium alloy fuel cladding within Pressurized Water Reactors (PWRs). DHC is caused by the accumulation of hydrogen atoms taken up by the metal, and the formation of brittle hydrides in the vicinity of crack tips. The rate of crack growth is limited by the rate of hydrogen diffusion to the crack, which can be modelled by solving a stress driven diffusion equation that incorporates the elastic interaction between defects. This of interest in the present work. The elastic interaction is calculated by combining defect forces determined through Density Functional Theory (DFT) simulations, and an exact solution for the anisotropic elastic field of an edge dislocation in Zr. making it possible to determine the interaction energy without the need to simulate directly a hydrogen atom in the presence of a crack or dislocation, which is computationally prohibitive with DFT. The result of the elastic interaction energy calculations can be utilised to determine the segregation of hydrogen to a crack tip for varying crack tip geometries, and in the presence of other crystal defects. This is done by implementing a diffusion equation for hydrogen within a discrete dislocation dynamics simulation. In the present work a model has been developed to demonstrate the effect of a single dislocation on hydrogen diffusion to create a Cottrell atmosphere.

  4. Oxidation behaviour of Zr-Ce alloys. Kinetic and microstructure aspects

    International Nuclear Information System (INIS)

    Rouillon, Ludovic

    1996-01-01

    As Zircaloy alloys are used for fuel rods in pressurized water nuclear reactors, this research thesis aims at studying and improving corrosion resistance of zirconium alloys while maintaining their mechanical properties. It more precisely deals with the kinetic and microstructure aspects of the external corrosion of the cladding by the coolant. In the case of Zircaloys, this corrosion is characterized by a kinetic transition from an initially parabolic to a linear regime. This research aims at intervening on this transition by elaborating zirconium alloys containing an element which stabilizes zirconia, in this case cerium. After having reported a bibliographical study on sheath oxidation, on parameters which influence sheath oxidation kinetics, on zirconia stabilization by doping elements, on the interest of lanthanide oxides, the author reports a feasibility study on the use of cerium (choice and preparation, sintered ceramic characterization, annealing of stabilized zirconia), reports a metallurgical study of Zr-Ce alloys, reports the study of the oxidation behaviour of these alloys (in autoclave, in presence of oxygen, under oxygen and then water) and the characterization of the microstructures of the oxide layers. He finally discusses the relationship between microstructure and oxidation kinetics, the role of cerium in the oxidation process, and the role of water in the oxidation process [fr

  5. Laser-Based Additive Manufacturing of Zirconium

    Directory of Open Access Journals (Sweden)

    Himanshu Sahasrabudhe

    2018-03-01

    Full Text Available Additive manufacturing of zirconium is attempted using commercial Laser Engineered Net Shaping (LENSTM technique. A LENSTM-based approach towards processing coatings and bulk parts of zirconium, a reactive metal, aims to minimize the inconvenience of traditional metallurgical practices of handling and processing zirconium-based parts that are particularly suited to small volumes and one-of-a-kind parts. This is a single-step manufacturing approach for obtaining near net shape fabrication of components. In the current research, Zr metal powder was processed in the form of coating on Ti6Al4V alloy substrate. Scanning electron microscopy (SEM and energy dispersive spectroscopy (EDS as well as phase analysis via X-ray diffraction (XRD were studied on these coatings. In addition to coatings, bulk parts were also fabricated using LENS™ from Zr metal powders, and measured part accuracy.

  6. Dielectric properties of zirconium dioxide-based ceramics

    International Nuclear Information System (INIS)

    Vladimirova, O.S.; Gruzdev, A.I.; Koposova, Z.L.; Lyutsareva, L.A.

    1985-01-01

    This paper studies the dielectric properties of materials based on stabilized zirconium dioxide with Co 3 O 4 additions possessing a high temperature-coefficient of resistance. These materials are promising for manufacturing resistance temperature gages that work under an oxidizing atmosphere at 370-1270 degrees K. The obtained results indicate the possibility of developing temperature gases possessing highsensitivity from stabilized zirconium dioxide with Co 3 O 4 additions

  7. Delayed hydride cracking in zirconium alloys in pressure tube nuclear reactors. Final report of a coordinated research project 1998-2002

    International Nuclear Information System (INIS)

    2004-10-01

    This report describes all of the research work undertaken as part of the IAEA coordinated research project on hydrogen and hydride induced degradation of the mechanical and physical properties of zirconium based alloys, and includes a review of the state of the art in understanding crack propagation by Delayed Hydride Cracking (DHC), and details of the experimental procedures that have produced the most consistent set of DHC rates reported in an international round-robin exercise to this date. It was concluded that 1) the techniques for performing measurements of the rate of delayed hydride cracking in zirconium alloys have been transferred from the host laboratory to other countries; 2) by following a strict procedure, a very consistent set of values of crack velocity were obtained by both individual laboratories and between the different laboratories; 3) the results over a wide range of test temperatures from materials with various microstructures fitted into the current theoretical framework for delayed hydride cracking; 4) an inter-laboratory comparison of hydrogen analysis revealed the importance of calibration and led to improvements in measurement in the participating laboratories and 5) the success of the CRP in achieving its goals has led to the initiation of some national programmes

  8. Research and development of zirconium industry in China

    International Nuclear Information System (INIS)

    Liu Jianzhang; Tian Zhenye

    2001-01-01

    The development of uranium material for nuclear power and silicon material for information industry represents two revolutionary changes in the material field in 20-th century. The development of these kinds of materials not only brings about great revolution of technology in the material field, but also promotes the great advancement of the world economy. Zirconium or its alloy, as one of the most important material in atomic age, just as the same as foreign countries has been developed under promotion of nuclear submarine project in China, and building of civil nuclear power reactor then has been laid a solid foundation for zirconium industry and provide a broad market for zirconium material

  9. Electrochemical oxidation of zirconium alloys in pre-transition and post-transition kinetic regimes at corrosion in electrolyte solutions

    International Nuclear Information System (INIS)

    Barkov, A.A.; Shavshin, V.M.

    1986-01-01

    With the aim of investigation on oxidation of zirconium alloys (Zr+2.5% Nb) the critical thickness of beginning of spalling of froming oxide films in HCl and NHO 3 aqueous solutions was evaluated by coulometry with accelerated procedure. Some variants of predeposition of modificated oxide coatings are proposed increase pre-transition regime time and to decrease corrosion during post-transition regime. Increase in agressivity of solutions (addition of 1 vol.% HF) and UV irradiation are found to increase 3-4 times pre-transition period

  10. Corrosion resistance and biocompatibility of zirconium oxynitride thin film growth by RF sputtering

    Energy Technology Data Exchange (ETDEWEB)

    Cubillos, G. I.; Olaya, J. J.; Clavijo, D.; Alfonso, J. E. [Universidad Nacional de Colombia, Carrera 45 No. 26-85, AA 14490 Bogota D. C. (Colombia); Bethencourt, M., E-mail: jealfonsoo@unal.edu.co [Universidad de Cadiz, Centro Andaluz de Ciencia y Tecnologia Marinas, Departamento de Ciencia de los Materiales e Ingenieria Metalurgica y Quimica Inorganica, Av. Republica de Saharaui, Puerto Real, E-11510 Cadiz (Spain)

    2012-07-01

    Thin films of zirconium oxynitride were grown on common glass, silicon (100) and stainless steel 316 L substrates using the reactive RF magnetron sputtering technique. The films were analyzed through structural, morphological and biocompatibility studies. The structural analysis was carried out using X-ray diffraction (XRD), and the morphological analysis was carried out using scanning electron microscopy (Sem) and atomic force microscopy (AFM). These studies were done as a function of growth parameters, such as power applied to the target, substrate temperature, and flow ratios. The corrosion resistance studies were made on samples of stainless steel 316 L coated and uncoated with Zr{sub x}N{sub y}O films, through of polarization curves. The studies of biocompatibility were carried out on zirconium oxynitride films deposited on stainless steel 316 L through proliferation and cellular adhesion. The XRD analysis shows that films deposited at 623 K, with a flow ratio {Phi}N{sub 2}/{Phi}O{sub 2} of 1.25 and a total deposit time of 30 minutes grew preferentially oriented along the (111) plane of the zirconium oxynitride monoclinic phase. The Sem analyses showed that the films grew homogeneously, and the AFM studies indicated that the average rugosity of the film was 5.9 nm and the average particle size was 150 nm. The analysis of the corrosion resistant, shows that the stainless steel coated with the film was increased a factor 10. Finally; through the analysis of the biocompatibility we established that the films have a better surface than the substrate (stainless steel 316 L) in terms of the adhesion and proliferation of bone cells. (Author)

  11. Peculiarities of formation of zirconium aluminides in hydride cycle mode

    International Nuclear Information System (INIS)

    Muradyan, G.N.

    2016-01-01

    The zirconium aluminides are promising structural materials in aerospace, mechanical engineering, chemical industry, etc. They are promising for manufacturing of heat-resistant wires, that will improve the reliability and efficiency of electrical networks. In the present work, the results of study of zirconium aluminides formation in the Hydride Cycle (HC) mode, developed in the Laboratory of high-temperature synthesis of the Institute of Chemical Physics of NAS RA, are described. The formation of zirconium aluminides in HC proceeded according to the reaction xZrH_2+(1-x)Al → alloy Zr_xAl(1-x)+H_2↑. The samples were certified using: chemical analysis to determine the content of hydrogen (pyrolysis method); differential thermal analysis (DTA, derivatograph Q-1500, T_heating = 1000°C, rate 20°C/min); X-ray analysis (XRD, diffractometer DRON-0.5). The influences of the ratio of powders ZrH_2/Al in the reaction mixture, compacting pressure, temperature and heating velocity on the characteristics of the synthesized aluminides were determined. In HC, the solid solutions of Al in Zr, single phase ZrAl_2 and ZrAl_3 aluminides and Zr_3AlH_4.49 hydride were synthesized. Formation of aluminides in HC mode took place by the solid-phase mechanism, without melting of aluminum. During processing, the heating of the initial charge up to 540°C resulted in the decomposition of zirconium hydride (ZrH_2) to HCC ZrH_1.5, that interacted with aluminum at 630°C forming FCC alumohydride of zirconium. Further increase of the temperature up to 800°C led to complete decomposition of the formed alumohydride of zirconium. The final formation of the zirconium aluminide occurred at 1000-1100°C in the end of HC process. Conclusion: in the synthesis of zirconium aluminides, the HC mode has several significant advantages over the conventional modes: lower operating temperatures (1000°C instead of 1800°C); shorter duration (1.5-2 hours instead of tens of hours); the availability of

  12. Corrosion-electrochemical behaviour and mechanical properties ofaluminium alloy-321, alloyed by barium

    International Nuclear Information System (INIS)

    Ganiev, I.; Mukhiddinov, G.N.; Kargapolova, T.V.; Mirsaidov, U.

    1995-01-01

    The purpose of present work is studying of influence of barium additionson electrochemical corrosion of casting aluminium-copper alloy Al-321,containing as base alloying components copper, chromium, manganese, titanium,zirconium, cadmium

  13. Preparation and certification of certified reference materials JAERI-Z21, Z22 and Z23 for analysis of zirconium and its alloys

    International Nuclear Information System (INIS)

    Takashima, Kyoichiro

    1991-03-01

    The Sub-Committee on Chemical Analysis of Nuclear Materials was organized in April 1987, under the Committee on Analytical Chemistry of Nuclear Fuels and Reactor Materials, JAERI, for renewal of certified reference materials of zirconium base alloys and zirconium metal. Collaborative analysis was carried out among ten participating laboratories for the certification of the JAERI CRMs Z21 to Z23. As a results of the collaborative works, the certified values for sixteen elements (Sn, Fe, Ni, Cr, Hf, Al, Si, Co, Cu, Ti, Mn, Pb, U, Cd, B and W) in the CRMs were given. In this report, preparation of raw materials, homogeneity test, chemical analysis for certification by collaborative works during April 1987 to March 1990 are described. (author)

  14. Microstructural characterization of Zr1Nb alloy after hot rolling

    Energy Technology Data Exchange (ETDEWEB)

    Souza, A.C. [Universidade Estadual do Mato Grosso do Sul (UEMS), MS (Brazil); Rossi, J.L.; Martinez, L.G.; Mucsi, C.S. [Instituto de Pesquisas Energéticas e Nucleares (IPEN/CNEN-SP), São Paulo, SP (Brazil); Tsakiropoulos, P. [University of Sheffield (United Kingdom); Ceoni, F.C.; Grandini, C.R. [Universidade Estadual Paulista Julio de Mesquita Filho (UNESP), SP (Brazil)

    2016-07-01

    Full text: The different research lines within the scope in engineering and materials science have developed new materials that can be used in different industrial sectors, such as, energy, health and transportation. For the nuclear industry, for example, the Zr alloys, are of great interest due to its good mechanical properties, excellent corrosion resistance and above all, the high permeability to thermal neutrons. In the health sector, the zirconium poses one of the lowest Young's modulus when compared to other metallic biomaterials, e.g., pure Zr is 68 GPa, bone mineral hydroxyapatite is 80 GPa, for Ti alloys is 90 GPa and above, for Nb is 105 GPa and stainless steels above 189 GPa. This is particularly important for implants in bones, whose elasticity modulus can reach 30 GPa and it is desirable an as close match as possible. However, the zirconium alloys, have great chemical affinity with oxygen and nitrogen. Moreover, oxides and nitrides may form during the melting process, heat treatment and hot rolling, changing the physic-chemical properties of the alloy. This experimental work shows the results of the evolution of the microstructure after hot rolling of the Zr1Nb alloy. It was possible to confirm the absence of formation of oxides and nitrides, thus confirming the of the experimental method of melting and hot rolling of the Zr1Nb alloy. (author)

  15. FUNDAMENTAL MECHANISMS OF CORROSION OF ADVANCED LIGHT WATER REACTOR FUEL CLADDING ALLOYS AT HIGH BURNUP

    International Nuclear Information System (INIS)

    Lott, Randy G.

    2003-01-01

    OAK (B204) The corrosion behavior of nuclear fuel cladding is a key factor limiting the performance of nuclear fuel elements, improved cladding alloys, which resist corrosion and radiation damage, will facilitate higher burnup core designs. The objective of this project is to understand the mechanisms by which alloy composition, heat treatment and microstructure affect corrosion rate. This knowledge can be used to predict the behavior of existing alloys outside the current experience base (for example, at high burn-up) and predict the effects of changes in operation conditions on zirconium alloy behavior. Zirconium alloys corrode by the formation f a highly adherent protective oxide layer. The working hypothesis of this project is that alloy composition, microstructure and heat treatment affect corrosion rates through their effect on the protective oxide structure and ion transport properties. The experimental task in this project is to identify these differences and understand how they affect corrosion behavior. To do this, several microstructural examination techniques including transmission electron microscope (TEM), electrochemical impedance spectroscopy (EIS) and a selection of fluorescence and diffraction techniques using synchrotron radiation at the Advanced Photon Source (APS) were employed

  16. Development of new zirconium based alloys for burn-up extension of light water reactor fuels, (1)

    International Nuclear Information System (INIS)

    Isobe, Takeshi; Matsuo, Yutaka

    1992-01-01

    Steam corrosion tests and tensile were conducted to investigate the effects of alloying elements such as Sn, Nb, Fe, Cr, Mo and V, and the mechanical properties of Nb-containing Zr-base alloys. The corrosion resistance of Zr-base alloys in comparison to Zr'y-4 was significantly improved by the reduction of the Sn content by 0.5 wt% and by a small addition of Nb (about 0.05 to 0.2 wt%). However, the decrease in solute Sn atoms degraded mechanical properties. The increase of the total content of Fe and Cr from 0.3 to 0.7 wt% improved the mechanical properties without affecting the corrosion resistance. The decrease of the Fe/Cr ratio from 6.0 to 0.5 increased the corrosion resistance. Small addition of Mo and/or V resulted in a further improvement of mechanical properties. Based on these experiments, three Nb-containing Zr-base alloys with equivalent mechanical properties and superior corrosion resistance to Zr'y-4 were developed. (author)

  17. Expanded heat treatment to form residual compressive hoop stress on inner surface of zirconium alloy tubing

    International Nuclear Information System (INIS)

    Megata, Masao

    1997-01-01

    A specific heat treatment process that introduces hoop stress has been developed. This technique can produce zirconium alloy tubing with a residual compressive hoop stress near the inner surface by taking advantage of the mechanical anisotropy in hexagonal close-packed zirconium crystal. Since a crystal having its basal pole parallel to the tangential direction of the tubing is easier to exhibit plastic elongation under the hoop stress than that having its basal pole parallel to the radial direction, the plastic and elastic elongation can coexist under a certain set of temperature and hoop stress conditions. The mechanical anisotropy plays a role to extend the coexistent stress range. Thus, residual compressive hoop stress is formed at the inner surface where more plastic elongation occurs during the heat treatment. This process is referred to as expanded heat treatment. Since this is a fundamental crystallographic principle, it has various applications. The application to improve PCI/SCC (pellet cladding interaction/stress corrosion cracking) properties of water reactor fuel cladding is promising. Excellent results were obtained with laboratory-scale heat treatment and an out-reactor iodine SCC test. These results included an extension of the time to SCC failure. (author)

  18. Thin polycrystalline diamond films protecting zirconium alloys surfaces: From technology to layer analysis and application in nuclear facilities

    Energy Technology Data Exchange (ETDEWEB)

    Ashcheulov, P. [Institute of Physics, Academy of Sciences Czech Republic v.v.i, Na Slovance 2, CZ-182 21, Prague 8 (Czech Republic); Škoda, R.; Škarohlíd, J. [Czech Technical University in Prague, Faculty of Mechanical Engineering, Technická 4, Prague 6, CZ-160 07 (Czech Republic); Taylor, A.; Fekete, L.; Fendrych, F. [Institute of Physics, Academy of Sciences Czech Republic v.v.i, Na Slovance 2, CZ-182 21, Prague 8 (Czech Republic); Vega, R.; Shao, L. [Texas A& M University, Department of Nuclear Engineering TAMU-3133, College Station, TX TX 77843 (United States); Kalvoda, L.; Vratislav, S. [Faculty of Nuclear Science and Physical Engineering, Czech Technical University in Prague, Brehova 7, CZ-115 19, Prague 1 (Czech Republic); Cháb, V.; Horáková, K.; Kůsová, K.; Klimša, L.; Kopeček, J. [Institute of Physics, Academy of Sciences Czech Republic v.v.i, Na Slovance 2, CZ-182 21, Prague 8 (Czech Republic); Sajdl, P.; Macák, J. [University of Chemistry and Technology, Power Engineering Department, Technická 3, Prague 6, CZ-166 28 (Czech Republic); Johnson, S. [Nuclear Fuel Division, Westinghouse Electric Company, 5801 Bluff Road, Hopkins, SC 29209 (United States); Kratochvílová, I., E-mail: krat@fzu.cz [Institute of Physics, Academy of Sciences Czech Republic v.v.i, Na Slovance 2, CZ-182 21, Prague 8 (Czech Republic); Faculty of Nuclear Science and Physical Engineering, Czech Technical University in Prague, Brehova 7, CZ-115 19, Prague 1 (Czech Republic)

    2015-12-30

    Graphical abstract: - Highlights: • In this work we showed that films prepared by MW-LA-PECVD technology can be used as anticorrosion protective layer for Zircaloy2 nuclear fuel claddings at elevated temperatures (950 °C) when α phase of zirconium changes to β phase (more opened for oxygen/hydrogen diffusion). Quality of PCD films was examined by Raman spectroscopy, XPS, SEM, AFM and SIMS analysis. • The polycrystalline diamond films were of high quality - without defects and contaminations. After hot steam oxidation (950 °C) a high level of structural integrity of PCD layer was observed. Both sp{sup 2} and sp{sup 3} C phases were present in the protective PCD layer. Higher resistance and a lower degree of impedance dispersion was found in the hot steam oxidized PCD coated Zircaloy2 samples, which may suggest better protection of the Zircaloy2 surface. The PCD layer blocks the hydrogen diffusion into the Zircaloy2 surface thus protecting the material from degradation. • Hot steam oxidation tests confirmed that PCD coated Zircaloy2 surfaces were effectively protected against corrosion. Presented results demonstrate that the PCD anticorrosion protection can significantly prolong service life of Zircaloy2 nuclear fuel claddings in nuclear reactors even at elevated temperatures. - Abstract: Zirconium alloys can be effectively protected against corrosion by polycrystalline diamond (PCD) layers grown in microwave plasma enhanced linear antenna chemical vapor deposition apparatus. Standard and hot steam oxidized PCD layers grown on Zircaloy2 surfaces were examined and the specific impact of polycrystalline Zr substrate surface on PCD layer properties was investigated. It was found that the presence of the PCD coating blocks hydrogen diffusion into the Zircaloy2 surface and protects Zircaloy2 material from degradation. PCD anticorrosion protection of Zircaloy2 can significantly prolong life of Zircaloy2 material in nuclear reactors even at temperatures above Zr

  19. Analytical functions used for description of the plastic deformation process in Zirconium alloys WWER type fuel rod cladding under designed accident conditions

    International Nuclear Information System (INIS)

    Fedotov, A.

    2003-01-01

    The aim of this work was to improve the RAPTA-5 code as applied to the analysis of the thermomechanical behavior of the fuel rod cladding under designed accident conditions. The irreversible process thermodynamics methods were proposed to be used for the description of the plastic deformation process in zirconium alloys under accident conditions. Functions, which describe yielding stress dependence on plastic strain, strain rate and temperature may be successfully used in calculations. On the basis of the experiments made and the existent experimental data the dependence of yielding stress on plastic strain, strain rate, temperature and heating rate for E110 alloy was determined. In future the following research work shall be made: research of dynamic strain ageing in E635 alloy under different strain rates; research of strain rate influence on plastic strain in E635 alloy under test temperature higher than 873 K; research of deformation strengthening of E635 alloy under high temperatures; research of heating rate influence n phase transformation in E110 and E635 alloys

  20. Electrical resistivity of liquid Ag-Au alloy

    International Nuclear Information System (INIS)

    Anis Alam, M.; Tomak, M.

    1983-01-01

    Calculations of the dependence of the electrical resistivity in liquid Ag-Au binary alloy on composition are reported. The structure of the binary alloy is described as a hard-sphere system. A one-parameter local pseudopotential, which incorporates s-d hybridization effects phenomenologically, is employed in the resistivity calculation. A reasonable agreement with experimental trend is observed. (author)

  1. Investigation of strain heterogeneities by laser scanning extensometry in strain ageing materials: application to zirconium alloys

    Energy Technology Data Exchange (ETDEWEB)

    Graff, S.; Forest, S.; Strudel, J.L. [Centre des Materiaux / UMR 7633, Ecole des Mines de Paris / CNRS, BP 87, 91003 Evry (France); Dierke, H.; Neuhauser, H. [Institut fur Physik der Kondensierten Materie, 38106 Braunschweig (Germany); Prioul, C. [MSSMAT, Ecole Centrale Paris, Grande Voie des Vignes, 92295 Chatenay-Malabry (France); Bechade, J.L. [SRMA, CEA Saclay, 91191 Gif sur Yvette (France)

    2005-07-01

    Laser scanning extensometry was used to detect and characterize propagating plastic instabilities such as the Luders bands at the millimeter scale. Spatio-temporal plastic heterogeneities are due to either static or dynamic strain ageing (SSA and DSA) phenomena. Regarding zirconium alloys, different type of heterogeneities were observed: their features strongly depended on mechanical test conditions. In one case, they appeared to be non propagating but preserved along the stress-strain curve and were associated with SSA effects such as stress peaks after relaxation periods or after unloading steps with waiting times. In other case, they appeared as non propagating but were not associated with SSA effects. (authors)

  2. Investigation of strain heterogeneities by laser scanning extensometry in strain ageing materials: application to zirconium alloys

    International Nuclear Information System (INIS)

    Graff, S.; Forest, S.; Strudel, J.L.; Dierke, H.; Neuhauser, H.; Prioul, C.; Bechade, J.L.

    2005-01-01

    Laser scanning extensometry was used to detect and characterize propagating plastic instabilities such as the Luders bands at the millimeter scale. Spatio-temporal plastic heterogeneities are due to either static or dynamic strain ageing (SSA and DSA) phenomena. Regarding zirconium alloys, different type of heterogeneities were observed: their features strongly depended on mechanical test conditions. In one case, they appeared to be non propagating but preserved along the stress-strain curve and were associated with SSA effects such as stress peaks after relaxation periods or after unloading steps with waiting times. In other case, they appeared as non propagating but were not associated with SSA effects. (authors)

  3. Corrosion behavior of AZ91 magnesium alloy treated by plasma immersion ion implantation and deposition in artificial physiological fluids

    International Nuclear Information System (INIS)

    Liu Chenglong; Xin Yunchang; Tian Xiubo; Chu, Paul K.

    2007-01-01

    Due to the good biocompatibility and tensile yield strength, magnesium alloys are promising in degradable prosthetic implants. The objective of this study is to investigate the corrosion behavior of surgical AZ91 magnesium alloy treated by aluminum, zirconium, and titanium plasma immersion ion implantation and deposition (PIII and D) at 10 kV in artificial physiological fluids. The surface layers show a characteristic intermixed layer and the outer surface are mainly composed of aluminum, zirconium or titanium oxide with a lesser amount of magnesium oxide. Comparing the three sets of samples, aluminum PIII and D significantly shifts the open circuit potential (OCP) to a more positive potential and improves the corrosion resistance at OCP

  4. Corrosion and Creep Characteristics of the HANA-4 Alloy with the various Manufacturing Processes

    International Nuclear Information System (INIS)

    Kim, Hyun-Gil; Choi, Byung-Kwan; Park, Jeong-Yong; Jeong, Yong-Hwan

    2008-01-01

    Zirconium alloys have been used as a fuel cladding material for several decades, since these alloys have revealed a good corrosion resistance and mechanical properties in reactor operating conditions. The development of an advanced Zr-based alloy with an improved corrosion and creep resistance is necessary for the high burn-up operating conditions in PWRs. The alloying element effects of the Nb, Sn, Fe, Cr, Cu etc as well as an optimization of the manufacturing processes such as the reduction ratio and annealing temperatures have been studied to improve the corrosion and creep properties. A high Nb-containing Zr-based alloy named HANA-4 was designed at KAERI and its nominal composition is Zr-1.5Nb-0.4Sn-0.2Fe-0.1Cr in wt.%. For high Nb-containing Zr alloys, their corrosion resistance is very sensitive to their microstructural characteristics which are determined by a manufacturing process. In order to obtain the best manufacturing process for the HANA-4 alloy, various evaluations such as corrosion and creep tests, a microstructural analysis, and a texture analysis were performed on the HANA-4 alloy with various manufacturing processes

  5. Mitigation of harmful effects of welds in zirconium alloy components

    International Nuclear Information System (INIS)

    Coleman, C.E.; Doubt, G.L.; Fong, R.W.L.; Root, J.H.; Bowden, J.W.; Sagat, S.; Webster, R.T.

    1995-01-01

    Welding produces local residual tensile stresses and changes in texture in components made from zirconium alloys. In the heat-affected zone in tubes or plates, the basal plane normals are rotated into the plane of the component and perpendicular to the direction of the weld. Thin-walled Zircaloy-2 tubes containing an axial weld do not reach their full strength because they always fail prematurely in the weld when pressurized to failure in a fixed-end burst test. Reinforcing the weld by increasing its thickness by 25% moves the failure to the parent metal and improves the biaxial strength of the tube by 20 to 25% and increases the total elongation by 200 to 450%. In components made from Zr-2.5Nb, the texture in the heat-affected zone promotes delayed hydride cracking (DHC) driven by tensile residual stress. Although the texture is not much affected by heat-treatments below 630 o C and large grain interaction stresses remain as a result of mixed textures, macro-residual tensile stresses can be relieved by heat treatment to the point where the probability of cracking is very low. (author)

  6. Mitigation of harmful effects of welds in zirconium alloy components

    International Nuclear Information System (INIS)

    Coleman, C.E.; Doubt, G.L.; Fong, R.W.L.; Root, J.H.; Bowden, J.W.; Sagat, S.; Webster, R.T.

    1993-10-01

    Welding produces local residual tensile stresses and changes in texture in components made from zirconium alloys. In the heat-affected zone in tubes or plates, the basal plane normals are rotated into the plane of the component and perpendicular to the direction of the weld. Thin-walled zircaloy-2 tubes containing an axial weld do not reach their full strength, because they always fail prematurely in the weld when pressurised to failure in a fixed-end burst test. Reinforcing the weld by increasing its thickness by 25% moves the failure to the parent metal, improves the biaxial strength of the tube by 20 to 25%, and increases the total elongation by 200 to 450%. In components made from Zr-2.5Nb, the texture in the heat-affected zone promotes delayed hydride cracking (DHC) driven by tensile residual stress. Although the texture is not much affected by heat-treatments below 630 degrees celsius and large grain interaction stresses remain as a result of mixed textures, macro-residual tensile stresses can be relieved by heat-treatment to the point where the probability of cracking is very low

  7. Mitigation of harmful effects of welds in zirconium alloy components

    International Nuclear Information System (INIS)

    Coleman, C.E.; Doubt, G.L.; Fong, R.W.L.; Root, J.H.; Bowden, J.W.; Sagat, S.

    1994-01-01

    Welding produces local residual tensile stresses and changes in texture in components made from zirconium alloys. In the heat-affected zone in tubes or plates, the basal plane normals are rotated into the plane of the component and perpendicular to the direction of the weld. Thin-walled Zircaloy-2 tubes containing an axial weld do not reach their full strength because they always fail prematurely in the weld when pressurized to failure in a fixed-end burst test. Reinforcing the weld by increasing its thickness by 25% moves the failure to the parent metal and improves the biaxial strength of the tube by 20 to 25% and increases the total elongation by 200 to 450%. In components made from Zr-2.5Nb, the texture in the heat-affected zone promotes delayed hydride cracking (DHC) driven by tensile residual stress. Although the texture is not much affected by heat-treatments below 630 C and large grain interaction stresses remain as a result of mixed textures, macro-residual tensile stresses can be relieved by heat treatment to the point where the probability of cracking is very low

  8. Deformation of zirconium - niobium alloy E635 in sub-microsecond shock waves

    Science.gov (United States)

    Kazakov, D. N.; Kozelkov, O. E.; Mayorova, A. S.; Malyugina, A. S.; Mokrushin, S. S.; Pavlenko, A. V.

    2015-09-01

    Strength characteristics of zirconium - niobium alloy E635 were measured under shock - wave loading conditions at normal and elevated temperatures and results of these measurements are presented. Measurements were taken in conditions when samples were impacted by plane shock waves with the pressure up to 13 GPa and duration from ˜0.05 μs up to 1 μs. Free-surface velocity profiles were recorded with the help of VISAR and PDV laser Doppler velocimeters having nanosecond time resolution. Evolution of elastic precursors with samples thickness varying from 0.5 up to 8 mm is also considered. Measured attenuation of the elastic precursor was used to determine plastic strain rate behind the precursor front. Temperature effect on the value of dynamic elastic limit and spall strength at normal and elevated temperatures is studied. This work is implemented with the support of the State Atomic Energy Corporation "Rosatom" under State Contract H.4x.44.90.13.1111.

  9. Corrosion resistant alloy uses in the power industry

    International Nuclear Information System (INIS)

    Nickerson, J.L.; Hall, F.A.; Asphahani, A.I.

    1989-01-01

    Nickel-base alloys have been used as cost-effective measures in a variety of severely corrosive situations in pollution control units for coal-fired power plants. Cost effectiveness and practical answers to corrosion problems are illustrated (specifically the wallpaper concept/metallic lining technique). Numerous cases of successful use of HASTELLOY alloys in Flue Gas Desulfurization (FGD) systems and hazardous waste treatment incineration scrubber systems are listed. In this paper developments in nickel-base alloys and their use in FGD and other segments of the power industry are discussed. In the Ni-Cr-Mo-W alloy family, the C-22 alloy has the best resistance to localized corrosion in halide environments (chloride/fluoride-containing solutions). This alloy is also used effectively as a universal filler metal to weld less-resistant alloys were weld corrosion may be a problem. Field performance of this alloy in the power industry is described

  10. Impact of β- radiolysis and transient products on irradiation-enhanced corrosion of zirconium alloys

    International Nuclear Information System (INIS)

    Lemaignan, C.

    1992-01-01

    An analysis has been undertaken of the various cases of local enhancement of the corrosion rate of zirconium alloys under irradiation. It is observed that in most cases a strong emission of energetic β - is present leading to a local energy desorption rate higher than the core average. This suggests that the local transient radiolytic oxidising species produced in the coolant by the β - particles could contribute to corrosion enhancement, by increasing the local corrosion potential. This process is applicable to the local enhanced corrosion found in front of stainless steels structural parts, due to the contribution of Mn, in front of Pt inserts and Cu-rich cruds. It explains also the irradiation corrosion enhancement of Cu-rich Zr alloys. Enhanced corrosion around neutron absorbing material is explained similarly by pair production from conversion of high energy capture photons in the cladding, leading to energetic electrons. The same process was found to be active with other highly ionising species like α in Ni-rich alloys and fission products in homogeneous reactors. This mechanism, applicable for an explanation of localised irradiation-enhanced corrosion, is proposed to be extended to the reactor core, where the general enhancement of Zr-alloy corrosion under irradiation would be due to the general radiolysis. It suggests that care should be taken to avoid any source of β - emission or other ionising species in the reactor core that could give an increase of energy deposition rate for radiolysis. Also the corrosion testing conditions for the materials to be used in reactors have to be relevant to the radiolytic environments found in the reactor cores. (orig.)

  11. PCI resistant light water reactor fuel cladding

    International Nuclear Information System (INIS)

    Foster, J.P.; Sabol, G.P.

    1988-01-01

    A tubular nuclear fuel element cladding tube is described, the fuel element cladding tube forming the entire fuel element cladding and consisting of: a single continuous wall, the single continuous wall consisting of a single alloy selected from the group consisting of zirconium base alloys, A, B, C, D, and E; the single continuous wall characterized by a cold worked and stress relieved microstructure throughout; wherein the zirconium base alloy A contains 0.2 - 0.6 w/o Sn, 0.03 - 0.11 w/o sum of Fe and Cr, section 600 ppm O and section 1500 ppm total impurities; the zirconium base alloy B contains 0.1 - 0.6 w/oo Sn, 0.04 - 0.24 w/o Fe, 0.05 - 0.15 w/o Cr, section 0.08 w/o Ni, section 600 ppm O and section 1500 ppm total impurities; the zirconium base alloy C contains 1.2 - 1.7 w/o Sn, 0.04 - 0.24 w/o Fe, 0.05 - 0.15 w/o Cr, section 0.08 w/o Ni, section 600 ppm O, and section 1500 ppm total impurities; the zirconium base alloy D contains 0.15 - 0.6 w/o Sn, 0.15 - 0.5 w/o Fe, section 600 ppm O, and section 1500 ppm total impurities; and the zirconium base alloy E contains 0.4 - 0.6 w/o Sn, 0.1 - 0.3 w/o Fe, 0.03 - 0.07 w/o Ni, section 600 ppm O, and section 1500 ppm total impurities

  12. Extra spots in the electron diffraction patterns of neutron irradiated zirconium and its alloys

    International Nuclear Information System (INIS)

    Madden, P.K.

    1977-01-01

    Specimens of neutron irradiated zirconium and its alloys were examined in the transmission electron microscope. Groups of extra spots, often exhibiting four-fold symmetry, were observed in thin foil electron diffraction patterns of these specimens. The 'extra-spot' structure, like the expected black-dot/small scale dislocation loop neutron irradiated damage, is approximately 100 A in size. Its nature is uncertain. It may be related to irradiation damage or to some artefact introduced during specimen preparation. If it is the latter, then published irradiation damage defect size distributions and determined irradiation growth strains of other investigators, may require modification. The present inconclusive results indicate that extra-spot structure is likely to consist of oxide particles, but may correspond to hydride precipitation or decoration effects, or even, to electron beam effects. (author)

  13. Determination of plastic anisotropy of zirconium alloys cladding

    International Nuclear Information System (INIS)

    Yamshchikov, N.V.; Prasolov, P.F.; Shestak, V.E.

    1991-01-01

    Method for determining plastic anisotropy of zurconium alloy cladding is described. It is based on consideration of material as a combination of transversal crystallites with known distribution over orientations. Such approach enables to describe cladding resistance to plastic deformation at arbitrary stressed state, using the results of texture investigations and uniaxial tests of samples, cut out of claddings along three directions. Plastic anisotropy of fuel element claddings 9.15 and 13.6 mm in diameter up to several percents of plastic deformation is shown

  14. Correlation between zirconium oxide impedance and corrosion behavior of Zr-Nb-Sn-Fe-Cu alloys

    International Nuclear Information System (INIS)

    Park, Sang Yoon; Lee, Myung Ho; Choi, Byoung Kwon; Jeong, Yong Hwan; Jung, Youn Ho

    2001-01-01

    To evaluate the correlation of Zr oxide impedance and corrosion behavior of Zr-Nb-Sn-Fe-Cu alloys, the corrosion behavior of the alloys was tested in the autoclave containing 70 ppm LiOH solution at 360 .deg. C. The characteristics of the oxide on the alloys were investigated by using the electrochemical impedance spectrosocpy (EIS) method. The corrosion resistance of the alloys was evaluated from the corrosion rate determined as a function of the concentration of Nb. The equivalent circuit of the oxide was composed on the base of the spectrum from EIS measurements on the oxide layers that had formed at pre-and post-transition regions on the curve of corrosion rate. By using the capacitance characteristics of the equivalent circuit, the thickness of impervious layer, it's electrical resistance and characteristics of space charge layer were evaluated. The corrosion characteristics of the Zr-Nb-Sn-Fe-Cu alloys were successfully explained by applying the EIS test results

  15. Oxidized zirconium on ceramic; Catastrophic coupling.

    Science.gov (United States)

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

    2017-02-01

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

  16. [Study on corrosion resistance of three non-noble porcelain alloys].

    Science.gov (United States)

    Wu, Zhikai; Xu, Sheng; Li, Wei; Teng, Jin; Li, Ning

    2011-10-01

    To study the electrochemical corrosion behavior of Co-Cr, Ni-Cr and Ni-Cr-Be based porcelain alloys in NaCl solution. Five samples of each alloy were made respectively, electric polarization curve of each alloy was obtained using potentiodynamic polarization technique. Self-corrosion potential (E(corr)), self-corrosion current density (I(corr), passive region and transpassivation potential were tested. Microstructure and constituent was examined using scanning electron microscopy and energy dispersive spectroscopy. Co-Cr alloy possessed the most desirable corrosion resistance because of its integrated, homogeneous and compact passive film. The poor compactness of Ni-Cr alloy's passive film decreased its corrosion resistance. Ni-Cr-Be alloy exhibited the worst corrosion resistance due to the Cr and Mo depleted Ni-Be eutectic phases in the alloy. Taking biological security into consideration, it is necessary to avoid the application of porcelain alloys with Be element. Co-Cr alloy with better biocompatibility possesses much broader prospect in the field of dental restoration.

  17. Manufacturing and performance tests of in-pile creep measuring machine of zirconium alloys

    International Nuclear Information System (INIS)

    Choi, Y.; Kim, B. G.; Kang, Y. H.

    2000-01-01

    A mock-up of the in-pile creep test machine of zirconium alloys for HANARO was designed and manufactured, which performance tests were carried. The dimension of the in-pile creep machine is 55 mm in diameter and 700 mm in length for HANARO, respectively. Load is transferred to specimen by through the working mechanisms in which the contraction of bellows by gas pressure moves a yoke and an upper grip connected to a specimen, simultaneously. It was observed that the extension of the specimen mounted in grips was transferred to a linear voltage differential transformer perfectly by a yoke and a push rod in a bearing. The displacement of specimen with applied pressure was determined with the LVDT and a pressure gauge, respectively. Resultant stress-strain behaviors of the specimen was determined by the displacement-applied gas pressure curve, which showed similar values obtained with a standard tensile test machine

  18. Electrical resistivity of liquid noble metal alloys

    International Nuclear Information System (INIS)

    Anis Alam, M.; Tomak, M.

    1983-08-01

    Calculations of the dependence of the electrical resistivity in liquid Ag-Au, Cu-Ag, Cu-Au binary alloys on composition are reported. The structure of the binary alloy is described as a hard sphere system. A one-parameter local pseudopotential, which incorporates s-d hybridization effects phenomenologically, is employed in the resistivity calculation. A reasonable agreement with experimental trends is observed in cases where experimental information is available. (author)

  19. Neutron resistant irradiation alloy and usage thereof

    International Nuclear Information System (INIS)

    Okada, Osamu; Nakata, Kiyotomo; Kato, Takahiko.

    1997-01-01

    A neutron irradiation embrittlement-resistant alloy comprising a Ti alloy having an average grain size of 2μm or smaller and containing from 30 to 40wt% of Al is subjected to powder solidification and then to isothermal forging at a forging rate of from 50 to 80% at a temperature range of from 1150 to 1500K. Namely, since the Ti-Al type alloy comprises from 30 to 30wt% of Al, optionally, from 1 to 6% of Mn, from 0.1 to 0.5% of Si, from 4 to 16% of V and the balance of Ti, it has excellent specific strength, high durable temperature and excellent neutron irradiation resistance, and has ductility required as structural materials. Accordingly, if the Ti-Al type alloy excellent in embrittlement resistance to neutron irradiation dimensional stability of materials is applied to constitutional parts of a reactor core of a nuclear reactor and a thermonuclear reactor to be exposed under neutron irradiation, high reliability is provided and the amount of activated materials is reduced by improving the working life of the materials. (N.H.)

  20. Determination of trace amounts of cadmium in zirconium and its alloys by graphite furnace AAS

    International Nuclear Information System (INIS)

    Takashima, Kyoichiro; Toida, Yukio

    1994-01-01

    Trace amount of cadmium in zirconium and its alloys was determined by graphite furnace atomic absorption spectrometry (GF-AAS) after ion exchange separation. A 2g chip sample was decomposed with 20ml of hydrofluoric acid (1+9) and a few drops of nitric acid. A trace amount of cadmium was separated from zirconium by strongly acidic cation-exchange resin (MCI GEL CK 08P) using 50ml of hydrochloric acid as an eluent. The solution was gently evaporated to dryness on an electric hot plate heater and under an infrared lamp. The residue was dissolved in 1ml of nitric acid (1+14) and diluted to 10ml in a volumetric glass flask with distilled water. Ten microliters of this solution was injected into a graphite furnace and then atomized at 2200degC for 4s in argon at a flow rate of 3.0l/min. Acids used in the analytical procedure were purified by azeotropic distillation and cation-exchange resin. The limit of determination (3σ BK ) for cadmium was 0.5ngCd/g and the relative standard deviation (RSD) at 1ngCd/g level was less than 20% for the GF-AAS. The accuracy of this technique was confirmed by NIST SRM 1643b (trace elements in water). (author)

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

    Directory of Open Access Journals (Sweden)

    Mushahary D

    2013-08-01

    Full Text Available Dolly Mushahary,1,2 Ragamouni Sravanthi,2 Yuncang Li,2 Mahesh J Kumar,1 Nemani Harishankar,4 Peter D Hodgson,1 Cuie Wen,3 Gopal Pande2 1Institute for Frontier Materials, Deakin University, Geelong, Australia; 2CSIR- Centre for Cellular and Molecular Biology, Hyderabad, India; 3Faculty of Engineering and Industrial Sciences, Swinburne University of Technology, Hawthorn, Australia; 4National Institute of Nutrition (ICMR, Tarnaka, Hyderabad, India Abstract: Development of new biodegradable implants and devices is necessary to meet the increasing needs of regenerative orthopedic procedures. An important consideration while formulating new implant materials is that they should physicochemically and biologically mimic bone-like properties. In earlier studies, we have developed and characterized magnesium based biodegradable alloys, in particular magnesium-zirconium (Mg-Zr alloys. Here we have reported the biological properties of four Mg-Zr alloys containing different quantities of strontium or calcium. The alloys were implanted in small cavities made in femur bones of New Zealand White rabbits, and the quantitative and qualitative assessments of newly induced bone tissue were carried out. A total of 30 experimental animals, three for each implant type, were studied, and bone induction was assessed by histological, immunohistochemical and radiological methods; cavities in the femurs with no implants and observed for the same period of time were kept as controls. Our results showed that Mg-Zr alloys containing appropriate quantities of strontium were more efficient in inducing good quality mineralized bone than other alloys. Our results have been discussed in the context of physicochemical and biological properties of the alloys, and they could be very useful in determining the nature of future generations of biodegradable orthopedic implants. Keywords: osteoblasts, bone mineralization, corrosion, osseointegration, surface energy, peri-implant

  2. Zirconium intermetallics and hydrogen uptake during corrosion

    International Nuclear Information System (INIS)

    Cox, B.

    1987-04-01

    The routes by which hydrogen can enter zirconium alloys containing second phase particles during corrosion are discussed. Both direct diffusion through the bulk of the oxide film, and migration through second phase particles that intersect the surface are considered. An examination of results for hydrogen uptake by zirconium alloys during the early stages of oxidation, when the oxide film is still coherent, suggests that for Zr, Zr-1%Cu and Zr-1%Fe the hydrogen enters by diffusing through the bulk ZrO 2 film, whereas for the Zircaloys the primary migration route may be through the intermetallics. The steps in the latter process are discussed and the evidence available on the properties of the intermetallics collated. A comparison of these data with results for hydrogen uptake by two series of ternary alloys (Zr-1%Nb - 1%X, Zr-1%Cu - 1%X) suggests that high hydrogen uptakes often correlate with intermetallics with high hydrogen solubilities and vice versa. The properties of Zr(Fe/Cr) 2+x intermetallics are examined in an attempt to understand the behaviour of the Zircaloys, and it is concluded that present data establishing composition and unit cell dimensions for such intermetallic particles are not of sufficient accuracy to permit a correlation

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

    Energy Technology Data Exchange (ETDEWEB)

    Bohlen, Jan, E-mail: jan.bohlen@hzg.de; Wendt, Joachim; Nienaber, Maria; Kainer, Karl Ulrich; Stutz, Lennart; Letzig, Dietmar

    2015-03-15

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

  4. The effect of Zr on the microstructure and properties of Ti-35Nb-XZr alloy

    Energy Technology Data Exchange (ETDEWEB)

    Málek, Jaroslav, E-mail: malek@ujp.cz [UJP PRAHA a.s., Nad Kamínkou 1345, 156 00 Prague-Zbraslav (Czech Republic); Czech Technical University in Prague, Faculty of Mechanical Engineering, Karlovo náměstí 13, 121 35 Praha 2 (Czech Republic); Hnilica, František, E-mail: hnilica@ujp.cz [UJP PRAHA a.s., Nad Kamínkou 1345, 156 00 Prague-Zbraslav (Czech Republic); Veselý, Jaroslav, E-mail: vesely@ujp.cz [UJP PRAHA a.s., Nad Kamínkou 1345, 156 00 Prague-Zbraslav (Czech Republic); Smola, Bohumil, E-mail: smola@met.mff.cuni.cz [Charles University in Prague, Faculty of Mathematics and Physics, KeKarlovu 5, 121 16 Prague 2 (Czech Republic); Kolařík, Kamil, E-mail: kamil.kolarik@email.cz [Czech Technical University in Prague, Faculty of Nuclear Sciences and Physical Engineering, Trojanova 339/13, 120 00 Praha 2 (Czech Republic); Fojt, Jaroslav, E-mail: jaroslav.fojt@vscht.cz [Institute of Chemical Technology, Technická 5, 166 28 Prague (Czech Republic); Vlach, Martin, E-mail: martin.vlach@mff.cuni.cz [Charles University in Prague, Faculty of Mathematics and Physics, KeKarlovu 5, 121 16 Prague 2 (Czech Republic); Kodetová, Veronika [Charles University in Prague, Faculty of Mathematics and Physics, KeKarlovu 5, 121 16 Prague 2 (Czech Republic)

    2016-10-15

    The demand for biomaterials with high strength, low modulus, excellent biocompatibility and good corrosion resistance has led to the development of new alloys. Zirconium is known as a biocompatible element that can be used for alloying in titanium alloys. The effect of Zr on the mechanical and structural properties of a titanium alloy is studied in this paper. Binary Ti-35Nb alloy has been alloyed with various amounts of Zr (2, 4, 6 or 8 wt%). The specimens were thermo-mechanically processed (hot forged, solution treated 850 °C/0.5 h/water quenched, cold swaged and finally aged (at 400 °C or 450 °C for various periods). Cold-swaged alloys possess tensile strength of about 800 MPa, along with a low Young's modulus (~50 GPa). The elongation of all the alloys is more than 12%. The hardness increased during 400 °C annealing up to 370 HV10. The addition of Zr stabilized the β-phase and supports recrystallization and recovery processes. Corrosion resistance was also increased by the addition of Zr.

  5. 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.

  6. Challenges in design of zirconium alloy reactor components

    International Nuclear Information System (INIS)

    Kakodkar, Anil; Sinha, R.K.

    1992-01-01

    Zirconium alloy components used in core-internal assemblies of heavy water reactors have to be designed under constraints imposed by need to have minimum mass, limitations of fabrication, welding and joining techniques with this material, and unique mechanisms for degradation of the operating performance of these components. These constraints manifest as challenges for design and development when the size, shape and dimensions of the components and assemblies are unconventional or untried, or when one is aiming for maximization of service life of these components under severe operating conditions. A number of such challenges were successfully met during the development of core-internal components and assemblies of Dhruva reactor. Some of the then untried ideas which were developed and successfully implemented include use of electron beam welding, cold forming of hemispherical ends of reentrant cans, and a large variety of rolled joints of innovative designs. This experience provided the foundation for taking up and successfully completing several tasks relating to coolant channels, liquid poison channels and sparger channels for PHWRs and test sections for the in-pile loops of Dhruva reactor. For life prediction and safety assessment of coolant channels of PHWRs some analytical tools, notably, a computer code for prediction of creep limited life of coolant channels has been developed. Some of the future challenges include the development of easily replaceable coolant channels and also large diameter coolant channels for Advanced Heavy Water Reactor, and development of solutions to overcome deterioration of service life of coolant channels due to hydriding. (author). 5 refs., 13 figs., 1 tab

  7. The chemical vapor deposition of zirconium carbide onto ceramic substrates

    International Nuclear Information System (INIS)

    Glass A, John Jr.; Palmisiano, Nick Jr.; Welsh R, Edward

    1999-01-01

    Zirconium carbide is an attractive ceramic material due to its unique properties such as high melting point, good thermal conductivity, and chemical resistance. The controlled preparation of zirconium carbide films of superstoichiometric, stoichiometric, and substoichiometric compositions has been achieved utilizing zirconium tetrachloride and methane precursor gases in an atmospheric pressure high temperature chemical vapor deposition system

  8. In situ monitored in-pile creep testing of zirconium alloys

    Science.gov (United States)

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

    2014-01-01

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

  9. Corrosion resistance of amorphous NiCrZr and NiCrMoZr alloys

    International Nuclear Information System (INIS)

    Naka, M.; Miyake, M.; Okamoto, I.

    1987-01-01

    One of the authors has reported that the corrosion resistance of chromium containing amorphous alloys is extremely improved by alloying phosphorus among metalloids. Two factors operate for the improvement of corrosion resistance of the amorphous alloys. First, phosphorus serves for the rapid formation of protective passive film. Second, the compositional and structural homogeneity in amorphous state also account for the formation of protective film. The latter factor has been clearly seen in the high corrosion resistance of CoCrMoZr and CoCrWZr alloys without metalloids. In order to clarify the separately two factors in the corrosion resistance of amorphous alloys, the corrosion resistance of amorphous alloys without metalloids has to be further investigated. This paper also deals with the corrosion resistance and electrochemical behavior of NiCrZr and NiCrMoZr alloys in 1N HCl, and compare them with the corrosion behavior of the crystalline alloys containing the same composition as that of the amorphous alloys

  10. Deformation of zirconium – niobium alloy E635 in sub-microsecond shock waves

    Directory of Open Access Journals (Sweden)

    Kazakov D.N.

    2015-01-01

    Full Text Available Strength characteristics of zirconium - niobium alloy E635 were measured under shock - wave loading conditions at normal and elevated temperatures and results of these measurements are presented. Measurements were taken in conditions when samples were impacted by plane shock waves with the pressure up to 13 GPa and duration from ∼0.05 μs up to 1 μs. Free-surface velocity profiles were recorded with the help of VISAR and PDV laser Doppler velocimeters having nanosecond time resolution. Evolution of elastic precursors with samples thickness varying from 0.5 up to 8 mm is also considered. Measured attenuation of the elastic precursor was used to determine plastic strain rate behind the precursor front. Temperature effect on the value of dynamic elastic limit and spall strength at normal and elevated temperatures is studied. This work is implemented with the support of the State Atomic Energy Corporation “Rosatom” under State Contract H.4x.44.90.13.1111.

  11. Corrosion resistance of aluminum-magnesium alloys in glacial acetic acid

    International Nuclear Information System (INIS)

    Zaitseva, L.V.; Romaniv, V.I.

    1984-01-01

    Vessels for the storage and conveyance of glacial acetic acid are produced from ADO and AD1 aluminum, which are distinguished by corrosion resistance, weldability and workability in the hot and cold conditions but have low tensile strength. Aluminum-magnesium alloys are stronger materials close in corrosion resistance to technical purity aluminum. An investigation was made of the basic alloying components on the corrosion resistance of these alloys in glacial acetic acid. Both the base metal and the weld joints were tested. With an increase in temperature the corrosion rate of all of the tested materials increases by tens of times. The metals with higher magnesium content show more pitting damage. The relationship of the corrosion resistance of the alloys to magnesium content is confirmed by the similar intensity of failure of the joint metal of all of the investigated alloys and by electrochemical investigations. The data shows that AMg3 alloy is close to technically pure ADO aluminum. However, the susceptibility of even this material to local corrosion eliminates the possibility of the use of aluminum-magnesium alloys as reliable constructional materials in glacial acetic acid

  12. Contribution to the identification of the processes kinetically limiting of the zirconium alloys oxidation; characterization of the oxide films formed at high temperature by solids electrochemistry

    International Nuclear Information System (INIS)

    Vermoyal, J.J.

    2000-06-01

    The corrosion behavior of zirconium alloys used for cladding tubes has been extensively studied under several oxidation conditions (temperature, steam, dry air, oxygen...) in order to clarify the mechanism(s) of oxide growth and breakdown. Oxidation rate is generally assumed to be controlled by oxygen diffusion inwards the oxide layer. Nevertheless, several experimental facts, such as acceleration or inhibition of corrosion rate in coupling conditions, suggest that electrochemical processes are involved as a rate determining step. This work is an attempt to shed light about the rate-limiting-mechanism of two zirconium alloys oxidation: Zircaloy-4 (Zy-4) and Zr-Nb(1%)O(0,13%). Impedance spectroscopy characterizations of oxide films formed in high temperature water and studied in gaseous atmosphere clearly show the difference of electrical properties between the two alloys. The in situ electrochemical and thermogravimetric investigations in gaseous medium, and the polarization effects on oxidation and hydridation of Zr alloys in PWRs conditions indicate that oxygen diffusion can be considered as the limiting kinetic step for Zy-4 oxidation. On the contrary, the acceleration of oxide growth on Zr-Nb(1%)O(0,13%) under anodic polarization in PWRs conditions (360 deg C) suggests that either the electronic conductivity in the oxide or an interfacial process at least partially control the oxidation rate. Catalytic effects observed in gaseous medium when noble metals increase the oxygen reduction rate would tend to corroborate the oxidation control of this alloy by an interfacial mechanism. An electrochemical description and a heterogeneous kinetics approach based on a diffusion-interfacial process as rate determining step are then proposed. (author)

  13. Study of the microstructural and mechanical properties of titanium-niobium-zirconium based alloys processed with hydrogen and powder metallurgy for use in dental implants

    International Nuclear Information System (INIS)

    Duvaizem, Jose Helio

    2009-01-01

    Hydrogen has been used as pulverization agent in alloys based on rare earth and transition metals due to its extremely high diffusion rate even on low temperatures. Such materials are used on hydrogen storage dispositives, generation of electricity or magnetic fields, and are produced by a process which the first step is the transformation of the alloy in fine powder by miling. Besides those, hydrogenium is also being used to obtain alloys based on titanium - niobium - zirconium in the pulverization. Powder metallurgy is utilized on the production of these alloys, making it possible to obtain structures with porous surface as result, requirement for its application as biomaterials. Other advantages of powder metallurgy usage include better surface finish and better microstructural homogeneity. In this work samples were prepared in the Ti-13Nb-13Zr composition. The hydrogenation was performed at 700 degree C, 600 degree C, and 500 degree C for titanium, niobium and zirconium respectively. After hydrogenation, the milling stage was carried out on high energy planetary ball milling with 200rpm during 90 minutes, and also in conventional ball milling for 30 hours. Samples were pressed in uniaxial press, followed by isostatic cold press, and then sintered at 1150 degree C for 7-13 hours. Microstructural properties of the samples were characterized by scanning electron microscope (SEM), energy dispersive spectroscopy (EDS) and x-ray diffraction. Mechanical and structural properties determined were density, microhardness and moduli of elasticity. The sample sintered at 1150 degree C for 7h, hydrogenated using 10.000 mbar and produced by milling on high energy planetary ball milling presented the best mechanical properties and microstructural homogeneity. (author)

  14. Devoluming method and device for radioactive metal wastes containing zirconium alloy

    International Nuclear Information System (INIS)

    Komatsu, Masahiko; Wada, Ryutaro.

    1996-01-01

    The present invention concerns a method of sealing radioactive metal wastes in a capsule and compressing the capsule for devoluming treatment. The method comprises a step of carrying radioactive metal wastes into a sealed chamber having a capacity somewhat greater than that of the capsule, a deaerating step of sucking the air in the sealed chamber to attain a substantially vacuum state, a compression-devoluming step of compression-devoluming the capsule by reducing the volume of the sealed chamber and a transporting step of transporting the devolumed capsule from the sealed chamber. The sealed chamber to which the capsule incorporated with radioactive metal wastes containing a zirconium alloy is carried is then deaerated into a substantially vacuum state. Even if ignitable powdery dusts are generated from the radioactive metal wastes crushed by compression-devoluming of the capsule in the succeeding compression-devoluming step, since the air necessary for ignition is not present, ignition of the powdery dusts is prevented. Alternatively, since the inside of the sealed chamber is filled with an inert gas, ignition of the powdery dusts can effectively be prevented. (N.H.)

  15. A layman's guide to radiation-induced deformation processes in zirconium alloys

    International Nuclear Information System (INIS)

    Dutton, R.

    1990-07-01

    The fuel channel (comprising a pressure tube and a calandria tube fabricated from zirconium alloys) in a CANDU reactor undergoes shape changes because of radiation-induced deformation. This is a consequence of the microstructural modification arising from radiation damage produced by the fast-neutron flux. This report summarizes our current understanding of the physical processes responsible for the deformation. With the non-specialist reader in mind, the underlying mechanisms are described in a manner that avoids much of the associated technical terminology. Thus, the basic concepts of plasticity in a crystalline material are introduced and related to the various microstructural defects created during irradiation. In particular, the mechanisms of creep (a time-dependent strain activated by an applied stress) and growth (a time-dependent strain occurring in the absence of stress) are discussed in a non-technical language assisted by simple diagrams. Reference is made to both theoretical investigations (avoiding mathematical complexity) and experimental measurements. It is shown how the qualitative and quantitative knowledge can be used to derive a predictive model for reactor designers and operators. The current status of such a model is evaluated and suggestions for future improvements made

  16. Microstructural changes in zirconium alloy bar due to multi-roll straightening

    International Nuclear Information System (INIS)

    Gouraharidas; Acharya, Swaroop; Pratap, Y.; Chaube, R.K.; Kiran Kumar, I.; Ramana Rao, A.V.; Saibaba, N.

    2010-01-01

    Zirconium alloy bar is the input material for making of end plugs required for encapsulating the uranium di-oxide pellets in the fuel tubes. These bars are manufactured through extrusion followed by multi-pass swaging with intermediate and final vacuum annealing. The straightened and ground bars are subjected to 100% Ultrasonic testing and Eddy current testing to identify flaws and micro-porosity in the material, which could otherwise affect the integrity of fuel element. The defect standards at ultrasonic and eddy current inspection have been made more stringent, in view of the importance of fuel pin integrity during reactor operation. Consequently, many of the rods have shown eddy current indications greater than the defect standard. Detailed microstructural examination was carried out at each process step to identify the cause for these indications. Characteristic variation in the grain size and microstructure were noticed from surface to the centre of the material. Correlation between residual stresses and the eddy current signals was established. The extent of residual stresses could be controlled by adopting improvised straightening method at the final stage. This paper deals with the various trials carried out and the conclusions arrived at. (author)

  17. Improved Oxidation Resistance of Zirconium Resulting from Addition of Small Amounts of Copper; Amelioration de la Resistance a l'Oxydation du Zirconium par de Faibles Additions de Cuivre; Povyshenie soprotivlyaemosti k okisleniyu tsirkoniya posle dobavlenij nebol'shikh kolichestv medi; Aumento de la Resistencia del Circonio a la Oxidacion por Agregado de Pequenas Cantidades de Cobre

    Energy Technology Data Exchange (ETDEWEB)

    Loriers, H.; Darras, R.; Baque, P. [Centre d' Etudes Nucleaires de Saclay (France)

    1966-02-15

    The authors have tried to explain the process by which the addition of small amounts of copper improves the resistance of zirconium to oxidation by pressurized carbon dioxide between 600 and 700 Degree-Sign C. The resulting alloys, particularly that with 2.5% copper, seem to offer interesting possibilities as cladding or structural materials for the fuel elements of reactors of the heavy water-carbon dioxide type. The corrosion of zirconium by carbon dioxide amounts essentially to oxidation, in which two processes are involved: (1) the building up of a layer of monoclinic zirconia, through which oxygen diffuses by migration of the anionic vacancies: (2) the dissolution and diffusion of oxygen, which penetrates into the metal. The combination of gravimetric data and measurements of the thickness of the oxide layers permits evaluation of the two corresponding diffusion coefficients. This calculation has been extended, by analogy, to the case of binary zirconium-copper alloys with a concentration of copper less than or equal to 4% by weight. Comparison of the results obtained shows that the apparent diffusion coefficient of oxygen in the oxide layer that is formed decreases as the copper content of the alloy in question increases; as regards the apparent diffusion coefficient of the oxygen in the alloy, though this is only slightly modified at 600 Degree-Sign C, at 700 Degree-Sign C it too decreases when the copper content is increased to 2.5%. The improved oxidation resistance associated with slowing down of the diffusion processes results, in particular, in prolongation of parabolic-type kinetics whereas, in the case of non-alloyed zirconium or other alloys, cracking in the oxide layer rapidly leads to linear oxidation kinetics as a result of the phenomenon of peeling. (author) [French] Les auteurs ont cherche a interpreter l'amelioration apportee par de faibles additions de cuivre a la resistance a l'oxydation du zirconium par le gaz carbonique sous pression entre

  18. Amorphous filler metal foils for brazing zirconium grid plates

    International Nuclear Information System (INIS)

    Plyushchev, A.N.; Kalin, B.A.; Fedotov, V.T.; Sevryukov, O.N.; Mamedova, T.T.; Shestakov, E.F.; Timoshin, S.N.

    2001-01-01

    A new amorphous ribbon filler metal of Zr-5.5 Fe-2.5 Be-1.0 Nb-8.0 Cu-2.0 Sn-0.4 Cr (mass %) with the temperature of melting onset of 745-750 deg C is designed to braze spacer grids of zirconium base alloys. The brazing conditions (780-790 deg C, 40-45 s) are determined which provide minimal standing at temperatures above 700 deg C (∼ 1.5 min) for spacer grids. Mechanical tests show that tensile strength of brazed joints is 55-59 kgf what is twice that of analogous welded joints. In addition, the brazed joints exhibit high corrosion resistance when testing in a distilled steam-water mixture at a temperature of 350 deg C and 16.5 MPa pressure for 10000 h [ru

  19. CORROSION RESISTANCE OF DYNAMIC LOADED CAST ALLOY AS12

    Directory of Open Access Journals (Sweden)

    A. A. Andrushevich

    2017-01-01

    Full Text Available The assessment of influence of powder particles in the mode of super deep penetration (SDP on change of corrosion resistance of aluminum cast alloy AK12 is executed. The aluminum alloy reinforced by fiber zones with the reconstructed structure has the increased corrosion resistance.

  20. Review of theoretical conceptions on regimes of oxidation and hydrogen pickup in Zr-alloys

    International Nuclear Information System (INIS)

    Likhanskii, V.; Evdokimov, I.

    2008-01-01

    In this paper the following issues are presented: 1) Experimental observations published in the journals on corrosion regimes of zirconium alloys of various compositions both for ex-pile oxidation experiments and for in-pile operating conditions of the materials. Factors experimentally stated on the effect of alloying composition, microstructure and texture on the rate of uniform corrosion and susceptibility of alloys to nodular corrosion. 2) Phenomenological models existing in publications, which describe conditions of uniform and nodular corrosion for Zr-alloys of various composition and microstructures, effect of irradiation and oxidizing medium; 3) Experimental data and phenomenological models describing regimes of hydrogen absorption in zirconium alloys; 4) Examples of application of physical models in explaining regimes, peculiarities of oxidation and hydrogen pickup for zirconium claddings of various alloying composition and microstructure

  1. Imparting passivity to vapor deposited magnesium alloys

    Science.gov (United States)

    Wolfe, Ryan C.

    Magnesium has the lowest density of all structural metals. Utilization of low density materials is advantageous from a design standpoint, because lower weight translates into improved performance of engineered products (i.e., notebook computers are more portable, vehicles achieve better gas mileage, and aircraft can carry more payload). Despite their low density and high strength to weight ratio, however, the widespread implementation of magnesium alloys is currently hindered by their relatively poor corrosion resistance. The objective of this research dissertation is to develop a scientific basis for the creation of a corrosion resistant magnesium alloy. The corrosion resistance of magnesium alloys is affected by several interrelated factors. Among these are alloying, microstructure, impurities, galvanic corrosion effects, and service conditions, among others. Alloying and modification of the microstructure are primary approaches to controlling corrosion. Furthermore, nonequilibrium alloying of magnesium via physical vapor deposition allows for the formation of single-phase magnesium alloys with supersaturated concentrations of passivity-enhancing elements. The microstructure and surface morphology is also modifiable during physical vapor deposition through the variation of evaporation power, pressure, temperature, ion bombardment, and the source-to-substrate distance. Aluminum, titanium, yttrium, and zirconium were initially chosen as candidates likely to impart passivity on vapor deposited magnesium alloys. Prior to this research, alloys of this type have never before been produced, much less studied. All of these metals were observed to afford some degree of corrosion resistance to magnesium. Due to the especially promising results from nonequilibrium alloying of magnesium with yttrium and titanium, the ternary magnesium-yttrium-titanium system was investigated in depth. While all of the alloys are lustrous, surface morphology is observed under the scanning

  2. Grain size determination in zirconium alloys. Final report of a co-ordinated research programme, 1989-1992

    International Nuclear Information System (INIS)

    1995-04-01

    A research programme was planned as an exercise to establish procedures and evaluate the success of technology transfer. The first programme under this scheme was proposed by the IAEA on the research topic: grain size determination in zirconium alloys. The host laboratory was Siemens AG Erlangen, in Germany. The programme was supervised by experts selected from participating countries. This report contains the results of the work carried out under this programme. The grain size of Zircaloy, the measurement methods, distribution of grain size in the matrix and dependence of grain size on temperature time of annealing are discussed in this report. The report also includes some information on the organizational arrangements and discusses possibilities for future collaboration. 38 figs, 11 tabs

  3. Advanced Corrosion-Resistant Zr Alloys for High Burnup and Generation IV Application

    International Nuclear Information System (INIS)

    Jeong, Y. H.; Park, S. Y.; Lee, M. H.; Choi, B. K.; Baek, J. H.; Park, J. Y.; Kim, J. H.; Kim, H. G.; Jung, Y. H.; Bang, B. G.

    2006-08-01

    The systematic study was performed to develop the advanced corrosion-resistant Zr alloys for high burnup and Gen IV application. The corrosion behavior was significantly changed with the alloy composition and the corrosion environment. In general, the model alloys with a higher alloying elements showed a higher corrosion resistance. Among the model alloys tested in this study, Zr-10Cr-0.2Fe showed the best corrosion resistance regardless of the corrosion condition. The oxide on the higher corrosion-resistant alloy such as Zr-1.0Cr-0.2Fe consisted of mainly columnar grains, and it have a higher tetragonal phase stability. In comparison with other alloys being considered for the SCWR, the Zr alloys showed a lower corrosion rate than ferritic-martensitic steels. The results of this study imply that, at least from a corrosion standpoint, Zr alloys deserve consideration as potential cladding or structural materials in supercritical water cooled reactors

  4. Fluorometric determination of zirconium in minerals

    Science.gov (United States)

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

    1951-01-01

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

  5. Comparison of the crevice corrosion resistance of alloys 625 and 22

    International Nuclear Information System (INIS)

    Palmer, J.; Kehler, B.; Iloybare, G.O.; Scully, J.R.

    1999-01-01

    The Yucca Mountain Site Characterization Project is concerned with the corrosion resistance of candidate engineered waste package materials. A variety of waste package designs have been proposed for US and Canadian High Level Nuclear Waste Repositories. A common feature of each design is the possibility of utilizing a corrosion resistant material such as a nickel-based super alloy or titanium-based alloy. A suitable corrosion resistant material may provide (a) kinetic immunity if the combination of repository environmental conditions and alloy resistance assure both: (i) a passive condition with negligible chance of localized corrosion stabilization, as well as (ii) low enough passive dissolution rates to insure conventional corrosion allowance over geological times, (b) a second form of ''corrosion allowance,'' if it can be scientifically demonstrated that a mechanism for stifling (i.e., death) of localized corrosion propagation occurs well before waste canisters are penetrated, or (c) such a low probability of initiation and continued propagation that a tolerably low degree of penetration occurs. Unfortunately, a large database on the crevice corrosion properties of alloy 22 does not exist in comparison to alloy 625. Alloy screening tests in oxidizing acids containing FeCl3 indicate that alloy 22 is more resistant to crevice corrosion than 625 as indicated by critical pit and crevice temperatures. Differences in alloying element compositions as expressed by pitting resistance equivalency number calculations support these findings. However, these data only provide the relative ranking of these alloys in terms of crevice corrosion and do not answer the critical questions proposed above

  6. Corrosion resistance of metallic materials for use in nuclear fuel reprocessing

    International Nuclear Information System (INIS)

    Legry, J.P.; Pelras, M.; Turluer, G.

    1989-01-01

    This paper reviews the corrosion resistance properties required from metallic materials to be used in the various developments of the PUREX process for nuclear fuel reprocessing. Stainless steels, zirconium or titanium base alloys are considered for the various plant components, where nitric acid is the main electrolyte with differing acid and nitrate concentrations, temperature and oxidizing species. (author)

  7. Interrelationship between structure and corrosion behaviour of zirconium

    Energy Technology Data Exchange (ETDEWEB)

    Guenther, T [Bayer A.G., Leverkusen (Germany, F.R.)

    1979-05-01

    Due to plant failures caused by the break-down of zirconium grade 702 subjected to sulphuric acid the structure and corrosion behaviour of welded and as delivered specimens were tested for various heat treatments. It was shown by structure investigations and electron microprobe analysis that the corrosion behaviour of zirconium (in boiling 65 pct sulphuric acid) is strongly infuenced by the structure, which in its turn is dependent on the grade of purity and the prehistory of the material. Type, amount, and distribution of residual elements or precipitations caused by them are responsible for the corrosion resistance. This is valid particularly for the element iron. The plant failures mentioned here coincided with the examination results. Measures to improve the chemical resistance of pure zirconium subjected to extremely aggressive media were derived.

  8. Zirconium and hafnium

    Science.gov (United States)

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

    2017-12-19

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

  9. Laser Surface Alloying of Aluminum for Improving Acid Corrosion Resistance

    Science.gov (United States)

    Jiru, Woldetinsay Gutu; Sankar, Mamilla Ravi; Dixit, Uday Shanker

    2018-04-01

    In the present study, laser surface alloying of aluminum with magnesium, manganese, titanium and zinc, respectively, was carried out to improve acid corrosion resistance. Laser surface alloying was conducted using 1600 and 1800 W power source using CO2 laser. Acid corrosion resistance was tested by dipping the samples in a solution of 2.5% H2SO4 for 200 h. The weight loss due to acid corrosion was reduced by 55% for AlTi, 41% for AlMg alloy, 36% for AlZn and 22% for AlMn alloy. Laser surface alloyed samples offered greater corrosion resistance than the aluminum substrate. It was observed that localized pitting corrosion was the major factor to damage the surface when exposed for a long time. The hardness after laser surface alloying was increased by a factor of 8.7, 3.4, 2.7 and 2 by alloying with Mn, Mg, Ti and Zn, respectively. After corrosion test, hardness was reduced by 51% for AlTi sample, 40% for AlMg sample, 41.4% for AlMn sample and 33% for AlZn sample.

  10. Corrosion Resistance of Co-Cr-Mo Alloy Used in Dentistry

    Directory of Open Access Journals (Sweden)

    Łukaszczyk A.

    2015-04-01

    Full Text Available The presented paper studies the effect of the casting technology on the corrosion resistance of Co-Cr-Mo alloy. The investigations were conducted on a commercial alloy with the brand name ARGELOY N.P SPECIAL (Co-Cr-Mo produced by Argen as well as the same alloy melted and cast by the lost wax casting method performed by a dental technician. The corrosion behavior of the dental alloys in an artificial saliva was studied with the use of the following electrochemical techniques: open circuit potential and voltammetry. After the electrochemical tests, studies of the surface of the examined alloys were performed by means of a scanning electron microscope with an X-ray microanalyzer. The results of the electrochemical studies show that the dependence of the corrosion resistance on the microstructure associated with the recasting process is marginal. The results of the electrochemical studies of the considered alloy clearly point to their good corrosion resistance in the discussed environment.

  11. Modelling of stress corrosion cracking in zirconium alloys

    International Nuclear Information System (INIS)

    Fandeur, O.; Rouillon, L.; Pilvin, P.; Jacques, P.; Rebeyrolle, V.

    2001-01-01

    During normal and incidental operating conditions, PWR power plants must comply with the first safety requirement, which is to ensure that the cladding wall is sound. Indeed some severe power transients potentially induce Stress Corrosion Cracking (SCC) of the zirconium alloy clad, due to strong Pellet Cladding Interaction (PCI). Since, at present, the prevention of this risk has some consequences on the French reactors manoeuvrability, a better understanding and forecast of the clad damage related to SCC/PCI is needed. With this aim, power ramp tests are performed in experimental reactors to assess the fuel rod behaviour and evaluate PCI failure risks. To study in detail SCC mechanisms, additional laboratory experiments are carried out on non-irradiated and irradiated cladding tubes. Numerical simulations of these tests have been developed aiming, on the one hand, to evaluate mechanical state variables and, on the other hand, to study consistent mechanical parameters for describing stress corrosion clad failure. The main result of this simulation is the determination of the validity ranges of the stress intensity factor, which is frequently used to model SCC. This parameter appears to be valid only at the onset of crack growth, when crack length remains short. In addition, the role of plastic strain rate and plastic strain as controlling parameters of the SCC process has been analysed in detail using the above mechanical description of the crack tip mechanical fields. Finally, the numerical determination of the first-order parameter(s) in the crack propagation rate law is completed by the development of laboratory tests focused on these parameters. These tests aim to support experimentally the results of the FE simulation. (author)

  12. Zirconium cladding - the long way towards a mechanistic understanding of processing and performance

    International Nuclear Information System (INIS)

    Preuss, Michael

    2011-01-01

    Zirconium alloys are the material of choice to encapsulate nuclear fuel in light and heavy water-cooled reactors due to their low neutron absorption, excellent corrosion resistance and sufficient mechanical properties. Despite these advantageous physical and mechanical properties a more physically based understanding of microstructure and texture evolution during processing is highly desirable in order to improve our understanding of formability during thermomechanical processing and performance variability of cladding material. In addition, the purely empirical understanding of aqueous zirconium corrosion, hydrogen pick up, hydride precipitation as well as irradiation growth and creep limits the accuracy of life predictions and therefore the level of burnup that is obtained from current fuel assemblies. The presentation aims at giving examples of new research strategies that will enable the development of a new physical understanding of processing and performance aspects in zirconium cladding material, which is required to develop new predictive models. Particular emphasis will be placed on using novel research tools and large-scale research facilities such as neutron spallation and synchrotron radiation sources to undertake very detailed and often in-situ studies of deformation mechanisms and microstructure evolution as well as determining stress states in grain families, oxides and hydrides. The results will be presented in the view of how they might help us to improve our understanding and enable the development of better predictive models

  13. Investigation of nano-structured Zirconium oxide film on Ti6Al4V substrate to improve tribological properties prepared by PIII&D

    Energy Technology Data Exchange (ETDEWEB)

    Saleem, Sehrish [Department of Physics, Government College University, Lahore 54000 (Pakistan); Department of Physics and Materials Science, City University of Hong Kong, Tat Chee Avenue, Kowloon (Hong Kong); Ahmad, R., E-mail: ahriaz@gcu.edu.pk [Department of Physics, Government College University, Lahore 54000 (Pakistan); Centre for Advanced Studies in Physics (CASP), Government College University, Lahore 54000 (Pakistan); Ayub, R. [Centre for Advanced Studies in Physics (CASP), Government College University, Lahore 54000 (Pakistan); Ikhlaq, Uzma [Department of Physics, Government College University, Lahore 54000 (Pakistan); Jin, Weihong; Chu, Paul K. [Department of Physics and Materials Science, City University of Hong Kong, Tat Chee Avenue, Kowloon (Hong Kong)

    2017-02-01

    Highlights: • ZrO{sub 2} film was deposited on Ti6Al4V alloy using the plasma immersion ion implantation and deposition at various bias voltage. • The deposited film was characterized by XPS, AFM, Ellipometry, Nano-indentation and Pin-on disk machine. • A dense zirconium oxide film with the maximum thickness 108 nm was formed at maximum applied voltage. • The hardness and wear resistance of film is much higher as compared to the substrate. - Abstract: Plasma immersion ion implantation and deposition (PIII&D) is the most attractive and efficient technique used in the medical field to tailor materials for biomedical applications. In the present study zirconium oxide nano-structured thin films were deposited on surface of Ti6Al4V alloy for bias voltages of 15, 20 and 25 kV. The chemical composition, surface roughness and thickness of deposited films were characterized by the x-ray photoelectron spectroscopy (XPS), atomic force microscope (AFM) and ellipsometry respectively. The XPS results confirm the formation of a dense zirconium oxide film of the treated specimens. AFM results exhibit a smooth film with maximum roughness of about 8.4 nm is formed. The thickness of the film is increased with the increase in bias voltages and is maximum at 25 kV. The effect of bias voltages on wear characteristics was further investigated by pin-on-disk test. It is observed that the friction coefficient is reduced, whereas wear resistance is enhanced and it is found to be maximum at 25 kV compared to the other bias voltages. Nanohardness is improved up to twice compared to untreated specimen at the maximum bias voltage. Therefore, it is concluded that deposition of zirconium oxide using the PIII&D is produced a dense layer on the substrate surface, which can be used as a promising candidate for the improved tribological properties of Ti6Al4V.

  14. 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)

  15. Moessbauer spectrometry study and metallography of paramagnetic phases from zirconium-iron system

    International Nuclear Information System (INIS)

    Freitas Brandao Bittencourt, C. de.

    1976-01-01

    Binary alloys of zirconium with 3 to 23% of iron by weight, were made by diffusion at 875 0 C of iron onto thin plates of zirconium. Moessbauer spectroscopy and optic metallography indicated the phases Zr 2 Fe and Zr 4 Fe, the bulk of which probably formed during the diffusion. These phases were confirmed by electron probe microanalysis. Moessbauer spectra showed quadrupole doublets with the same hyperfine interaction parameters in both phases, but with clearly distinct asymmetries. (author)

  16. 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

  17. Lithium uptake and the accelerated corrosion of zirconium alloys

    International Nuclear Information System (INIS)

    Ramasubramanian, N.; Precoanin, N.; Ling, V.C.

    1989-01-01

    The corrosion of zirconium alloys in aqueous lithiated solutions is sensitive to the concentration of the alkali and the temperature. In concentrated solutions, >10 -1 M in lithium hydroxide (LiOH) (700-ppm lithium) and at temperatures >573 K, accelerated corrosion occurs at quite an early stage. Our investigations indicate that the accelerated corrosion is caused by the generation of porosity, rather than the dissolution of lithium, in the growing oxide. Specimens of standard Zircaloy-4 fuel cladding and Zr-2.5 wt% Nb pressure tube materials were corroded in lithium hydroxide solutions, 10 -3 to 1 M in concentration, at 589 K. Impedance measurements, polarizations in molten lithium nitrate-lithium hydroxide (LiNO 3 -LiOH) and scanning electron microscopy of the alloy-oxide interface indicated a high level of porosity, right from the initial stages, for oxide films grown in the concentrated solutions. The oxides, when analyzed by atomic absorption spectroscopy, revealed the presence of a few 100 ppm of lithium, too small to account for the accelerated corrosion by a mechanism of solid solution of lithium in zirconia. X-ray powder patterns of the oxides showed peaks for only monoclinic zirconia, but occasionally peaks for LiOH · H 2 O and LiOH were also observed. The counts for lithium, detected by secondary ion mass spectrometry, decreased when specimens cut from the same corroded samples were leached in nitric acid. It is concluded from these observations that a major part of lithium is physically held in the porous oxide. Lithium hydroxide is not completely dissociated in aqueous solutions; with increasing concentration and temperature, an increasingly larger proportion of the alkali remains undissociated. It is suggested that the accelerated corrosion in concentrated solutions is caused by the participation of the undissociated alkali in the reactions occurring on the surfaces of the zirconia crystallites. The undissociated LiOH and hydroxyl ions react at an

  18. Degradation of protective properties of oxide films on Zr-Nb alloys after neutron and ion irradiation

    International Nuclear Information System (INIS)

    Belykh, T.A.; Kruzhalov, A.V.; Neshov, F.G.; Matveev, A.V.; Perekhozhev, V.I.; Sinel'nikov, L.P.; Kozlov, A.V.; Kalachikov, V.E.

    2000-01-01

    Specimens of zirconium-niobium alloys with surface oxide films, which had been working in technological channels of boiling nuclear reactor and were irradiated with nitrogen (10 MeV) and carbon (9 MeV) ions, were investigated by the methods of electrochemical decoration and resistance measuring. The role of the radiation effect in the processes of nodular corrosion formation and development is studied [ru

  19. 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.)

  20. Comments on the Dutton-Puls model: Temperature and yield stress dependences of crack growth rate in zirconium alloys

    International Nuclear Information System (INIS)

    Kim, Young S.

    2010-01-01

    Research highlights: → This study shows first that temperature and yield stress dependences of crack growth rate in zirconium alloys can analytically be understood not by the Dutton-Puls model but by Kim's new DHC model. → It is demonstrated that the driving force for DHC is ΔC, not the stress gradient, which is the core of Kim's DHC model. → The Dutton-Puls model reveals the invalidity of Puls' claim that the crack tip solubility would increase to the cooling solvus. - Abstract: This work was prompted by the publication of Puls's recent papers claiming that the Dutton-Puls model is valid enough to explain the stress and temperature dependences of the crack growth rate (CGR) in zirconium alloys. The first version of the Dutton-Puls model shows that the CGR has positive dependences on the concentration difference ΔC, hydrogen diffusivity D H , and the yield strength, and a negative dependence on the applied stress intensity factor K I , which is one of its critical defects. Thus, the Dutton-Puls model claiming that the temperature dependence of CGR is determined by D H C H turns out to be incorrect. Given that ΔC is independent of the stress, it is evident that the driving force for DHC is ΔC, not the stress gradient, corroborating the validity of Kim's model. Furthermore, the predicted activation energy for CGR in a cold-worked Zr-2.5Nb tube disagrees with the measured one for the Zr-2.5Nb tube, showing that the Dutton-Puls model is too defective to explain the temperature dependence of CGR. It is demonstrated that the revised Dutton-Puls model also cannot explain the yield stress dependence of CGR.

  1. A new model for the in-reactor corrosion of zirconium alloys

    International Nuclear Information System (INIS)

    Cox, B.

    1997-01-01

    Previous models for the in-reactor corrosion of zirconium alloys have assumed that the mechanism is a completely solid-state diffusion process, determined by the growth and breakdown of the protective oxide film. In-reactor kinetics have been related to out-reactor kinetics with the oxide-metal interface temperature calculated from effects of heat flux. Recent experimental results have suggested that oxide dissolution and reprecipitation may be a major process leading to the formation of thick porous oxide films in-reactor. The model described here is based on the dissolution of primary recoil tracks in the oxide as the primary process distinguishing in-reactor from out-reactor corrosion. The consequences of such a model would be a very different microscopic morphology of in-reactor and out-reactor thick films, a significant irradiation effect on non-heat transfer surfaces, and a change in the kinetics of the overall process. This model should be equally applicable to PWR and BWR water chemistries because of the amphoteric nature of ZrO 2 , and the effects of LiOH should operate by an essentially identical mechanism. A reciprocal rate equation should fit these processes and with additive terms seems capable of accommodating all water chemistry effects, except for discontinuous processes such as nodular corrosion. (author). 60 refs, 6 figs

  2. Corrosion resistance of Fe-based amorphous alloys

    International Nuclear Information System (INIS)

    Botta, W.J.; Berger, J.E.; Kiminami, C.S.; Roche, V.; Nogueira, R.P.; Bolfarini, C.

    2014-01-01

    Highlights: ► We report corrosion properties of Fe-based amorphous alloys in different media. ► The Cr-containing alloys had corrosion resistance close to that of Pt in all media. ► The wide range of electrochemical stability is relevant in many industrial domains. -- Abstract: Fe-based amorphous alloys can be designed to present an attractive combination of properties with high corrosion resistance and high mechanical strength. Such properties are clearly adequate for their technological use as coatings, for example, in steel pipes. In this work, we studied the corrosion properties of amorphous ribbons of the following Fe-based compositions: Fe 66 B 30 Nb 4 , [(Fe 0.6 Co 0.4 ) 0.75 B 0.2 Si 0.05 ] 96 Nb 4 , [(Fe 0.7 Co 0.3 ) 0.75 B 0.2 Si 0.05 ] 96 Nb 4 , Fe 56 Cr 23 Ni 5.7 B 16 , Fe 53 Cr 22 Ni 5.6 B 19 and Fe 50 Cr 22 Ni 5.4 B 23 . The ribbons were obtained by rapid solidification using the melt-spinning process, and were characterized by X-ray diffraction (XRD), differential scanning calorimetry (DSC) and optical (OM) and scanning electron microscopy (SEM). The corrosion properties were evaluated by corrosion potential survey and potentiodynamic polarization. The Cr containing alloys, that is the FeCrNiB type of alloys, showed the best corrosion resistance properties with the formation of a stable passive film that ensured a very large passivation plateau

  3. Bulk metallic glasses and high entropy alloys for reprocessing applications

    International Nuclear Information System (INIS)

    Kamachi Mudali, U.; Jayaraj, J.

    2016-01-01

    Recent breakthroughs in materials engineering have generated complex alloys that retain a glassy state in bulk form (bulk metallic glasses or BMGs) via ingot casting. High corrosion resistance is expected for BMGs (amorphous) as they are free from defects associated with the crystalline state such as grain boundaries, dislocations and stacking faults. Compared with conventional alloys containing one or two principal elements, the recently developed HEAs are usually composed of five or more elements with equimolar or near equimolar elemental fractions, which forms single solid solution phase. These HEAs exhibit excellent microstructural stability with better mechanical, wear and corrosion resistance properties as they are essentially single phase. Reprocessing of spent fuel from the fast breeder reactor involves the use of high concentration of (11.5 M) nitric acid under boiling conditions for the dissolution of the fuel. Conventional AISI type 304LSS and nitric acid grade 304L stainless steel would undergo inter-granular corrosion under these conditions and cannot be used for the fabrication of dissolver vessel. Currently titanium is used and zirconium alloys are proposed for future dissolver applications. Thus searching for newer materials with higher corrosion resistance suggests metallic glasses and HEAs for critical components of the dissolver application. Several Zr-based glassy alloys with different microstructural states and Ni-Nb based glassy alloys and TiZrHfNbTa HEA were cast and characterized for microstructure and corrosion resistance in nitric acid medium. From these studies, factors such as the corrosive environment (nitric acid, chloride and fluoride), and the presence of passivating elements in the alloy were emphasized for better corrosion resistance of BMGs and HEA. Attempts were also made to prepare coatings of Zr-and Ni-based glassy alloys on 304LSS by laser based deposition technique and their corrosion properties were evaluated. (author)

  4. A quantitative phase field model for hydride precipitation in zirconium alloys: Part I. Development of quantitative free energy functional

    International Nuclear Information System (INIS)

    Shi, San-Qiang; Xiao, Zhihua

    2015-01-01

    A temperature dependent, quantitative free energy functional was developed for the modeling of hydride precipitation in zirconium alloys within a phase field scheme. The model takes into account crystallographic variants of hydrides, interfacial energy between hydride and matrix, interfacial energy between hydrides, elastoplastic hydride precipitation and interaction with externally applied stress. The model is fully quantitative in real time and real length scale, and simulation results were compared with limited experimental data available in the literature with a reasonable agreement. The work calls for experimental and/or theoretical investigations of some of the key material properties that are not yet available in the literature

  5. Emerging technologies for processing and utilization of zircon

    International Nuclear Information System (INIS)

    Rao, S.M.; Balkishan, B.

    2013-01-01

    Zirconium occupies a prominent place in today's industry due to diverse applications of its chemical compounds. About 95% of zirconium is consumed in the form of zircon, zirconia and zirconium chemicals. The balance 5% of zirconium is consumed as the metal and its alloys. Because of its attractive nuclear properties zirconium alloys with very low hafnium content are used in the nuclear industry. Bio-compatibility, corrosion resistance and mechanical strength of zirconium alloys make them suitable for the manufacture of human organs and medicinal and pharmaceutical equipment. Other commercial areas of application of zirconium compounds include foundry, refractory, abrasives, opacifiers, oxygen sensors, catalysis, advanced ceramics, gems, paint dryers, pigments etc. The talk outlines the technologies involved in the production of nuclear grade zirconium alloys, manufacture of zirconium metal powder, gem grade zirconia and zirconium compounds like oxy-chloride, basic sulphate and basic carbonate etc. (author)

  6. Influence of preliminary radiation-oxidizing treatment on the corrosion resistance of zirconium in conditions of action of ionizing radiation

    International Nuclear Information System (INIS)

    Garibov, A. A.; Aliyev, A. G.; Agayev, T. N.; Velibekova, G. Z.

    2004-01-01

    Today mainly water-cooled nuclear reactors predominate in atomic energetics. For safe work of nuclear reactors detection of accumulation process of explosives, formed during radiation and temperature influence on heat-carriers in contact with materials of nuclear reactors in normal and emergency regimes of work is of great importance. The main sources of molecular hydrogen formation in normal and emergency regimes are the processes of liquid and vaporous water in vapo metallic reaction [1-5]. At the result of these processes molecular hydrogen concentration in heat-carrier composition always exceeds theoretically expected concentration. One of the main ways to solve the problem of water-cooled reactors safety is detection of possibilities to raise material resistance of fuel elements and heat carrier to joint action of ionizing radiation and temperature. The second way is inhibition of radiation-catalytic activity of construction materials' surface during the process of water decomposition. It's been established, that one of the ways to raise resistance of zirconium materials to the influence of ionizing radiation is formation of thin oxide film on the surface of metals. In the given work the influence of preliminary oxidizing treatment of zirconium surface on its radiation-catalytic activity during the process of water decomposition. With this aim zirconium is exposed to preliminary influence of gamma-quantum in contact with hydrogen peroxide at different meanings of absorbed radiation dose

  7. Method to electrolytically precipitate metals onto zirconium objects

    International Nuclear Information System (INIS)

    Donaghy, R.E.

    1978-01-01

    Tubes and other formed bodies made of zirconium or zirconium alloys which serve to take up nuclear fuels, are plated by electrolytically depositing a metal film onto these in order to improve their mechanical and corrosion properties. The object is activated in a solution of ammonium bifluoride and sulphuric acid, whereby an electrically conducting solid and a loose layer is formed. This loose film is removed by using fluoboric acid or hydrofluoric silicic acid solution, ultrasonics, or strips of organic material (cotton, polyester, nylon). The plating of Cu, Ni, Cr is described in detail. The object is rinsed between the process steps with deionized water and finally degased at a temperature of 150-200 0 C. (IHOE) [de

  8. Modelling of Zirconium and Hafnium separation using continuous annular chromatography

    International Nuclear Information System (INIS)

    Moch-Setyadji; Endang Susiantini

    2014-01-01

    Nuclear degrees of zirconium in the form of a metal alloy is the main material for fuel cladding of NPP. Zirconium is also used as sheathing UO 2 kernel in the form of ZrC as a substitute of SiC in the fuel elements of High Temperature Reactor (HTR). Difficulty separating hafnium from zirconium because it has a lot of similarities in the chemical properties of Zr and Hf. Annular chromatography is a device that can be used for separating of zirconium and hafnium to obtain zirconium nuclear grade. Therefore, it is necessary to construct the mathematical modelling that can describe the separation of zirconium and hafnium in the annular chromatography containing anion resin dowex-1X8. The aim of research is to perform separation simulation by using the equilibrium model and mass transfer coefficient resulted from research. Zr and Hf feed used in this research were 26 and 1 g/l, respectively. Height of resin (L), angular velocity (ω) and the superficial flow rate (uz) was varied to determine the effect of each parameter on the separation of Zr and Hf. By using Kd and Dv values resulted previous research. Simulation results showed that zirconium and hafnium can be separated using a continuous annular chromatography with high resin (long bed) 50 cm, superficial flow rate of 0.001 cm/s, the rotation speed of 0.006 rad/min and 20 cm diameter annular. In these conditions the results obtained zirconium concentration of 10,303.226 g/m 3 and hafnium concentration of 12.324 g/m 3 (ppm). (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    Coffey, Greg W. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Meinhardt, Kerry D. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Joshi, Vineet V. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Pederson, Larry R. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Lavender, Curt A. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Burkes, Douglas [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2015-03-01

    The Fuel Fabrication Capability within the U.S. High Performance Research Reactor Conversion Program is funded through the National Nuclear Security Administration (NNSA) NA-26 (Office of Material Management and Minimization). An investigation was commissioned to determine the feasibility of using electroplating techniques to apply a coating of zirconium onto depleted uranium/molybdenum alloy (U-10Mo). Electroplating would provide an alternative method to the existing process of hot roll-bonding zirconium foil onto the U-10Mo fuel foil during the fabrication of fuel elements for high-performance research reactors. The objective of this research was to develop a reproducible and scalable plating process that will produce a uniform, 25 μm thick zirconium metal coating on U-10Mo foil. In previous work, Pacific Northwest National Laboratory (PNNL) established a molten salt electroplating apparatus and protocol to plate zirconium metal onto molybdenum foil (Coffey 2015). During this second year of the research, PNNL furthered this work by moving to the U-10Mo alloy system (90 percent uranium:10 percent molybdenum). The original plating apparatus was disassembled and re-assembled in a laboratory capable of handling low-level radioactive materials. Initially, the work followed the previous year’s approach, and the salt bath composition was targeted at the eutectic composition (LiF:NaF:ZrF4 = 26:37:37 mol%). Early results indicated that the formation of uranium fluoride compounds would be problematic. Other salt bath compositions were investigated in order to eliminate the uranium fluoride production (LiF:NaF = 61:39 mol% and LiF:NaF:KF = 46.5:11.5:42 mol% ). Zirconium metal was used as the crucible for the molten salt. Three plating methods were used—isopotential, galvano static, and pulsed plating. The molten salt method for zirconium metal application provided high-quality plating on molybdenum in PNNL’s previous work. A key advantage of this approach is that

  10. 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

  11. Chemical aspects of hydrogen ingress in zirconium and zircaloy pressure tubes: ageing management of Indian PHWR coolant channels - determination of hydrogen and deuterium

    International Nuclear Information System (INIS)

    Sayi, Y.S.; Shankaran, P.S.; Yadav, C.S.; Ramanjaneyulu, P.S.; Venugopal, V.; Ramakumar, K.L.; Chhapru, G.C.; Prasad, R.; Jain, H.C.; Sood, D.D.

    2009-02-01

    Pressurized heavy water reactors (PHWRs) use zirconium and zirconium based alloys as clad and coolant tubes since its beginning. The first ever zircaloy-2 pressure tube failure occurred in 1983 at Ontario Hydro's Pickering Unit 2 in Canada which necessitated a thorough examination of causes of such failure. The failure was attributed to massive hydriding at the failed spot of pressure tube. Continuous usage of zirconium alloys could result in their hydrogen and deuterium pick-up leading to hydrogen/ deuterium embrittlement. The life of the zircaloy coolant channels is dictated by hydrogen/deuterium content and hence ageing management of the pressure tubes is essential for ensuring their trouble-free usage. It is desirable to have a sound knowledge on the chemical aspects of zirconium and zirconium based alloys metallurgy, the mechanistic principles of hydrogen ingress into the pressure tubes during in reactor service, and identifying suitable analytical methodologies for precise and accurate determination of hydrogen in wafer thin sliver samples carved out from insides of pressure tubes without causing any structural damage so that it can continue to remain in service. This is desirable so that the ageing management does not result in cost-escalation. This report is divided in to three main parts. The first part deals with the chemical aspects of zirconium and zirconium based alloy metallurgy, the mechanism of hydrogen pick-up and hydride formation in zirconium matrix. The second part describes various methodologies and their limitations, available for hydrogen/deuterium determination. The third part deals in detail, about the extensive investigations carried out at Radioanalytical Chemistry Division (RACD) in Radiochemistry and Isotope Group for establishing an indigenously developed hot vacuum extraction system in combination with quadrupole mass spectrometry for precise determination of hydrogen and deuterium in wafer thin sliver sample of zircaloy. The

  12. Corrosion resistance improvement of titanium base alloys

    Directory of Open Access Journals (Sweden)

    Mihai V. Popa

    2010-01-01

    Full Text Available The corrosion resistance of the new Ti-6Al-4V-1Zr alloy in comparison with ternary Ti-6Al-4V alloy in Ringer-Brown solution and artificial Carter-Brugirard saliva of different pH values was studied. In Ringer-Brown solution, the new alloy presented an improvement of all electrochemical parameters due to the alloying with Zr; also, impedance spectra revealed better protective properties of its passive layer. In Carter-Brugirard artificial saliva, an increase of the passive film thickness was proved. Fluoride ions had a slight negative influence on the corrosion and ion release rates, without to affect the very good stability of the new Ti-6Al-4V-1Zr alloy.

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

    Science.gov (United States)

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

    2018-01-01

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

  14. Ferritic Alloys as Accident Tolerant Fuel Cladding Material for Light Water Reactors

    International Nuclear Information System (INIS)

    Rebak, Raul B.

    2014-01-01

    The objective of the GE project is to demonstrate that advanced steels such as iron-chromium-aluminum (FeCrAl) alloys could be used as accident tolerant fuel cladding material in commercial light water reactors. The GE project does not include fuel development. Current findings support the concept that a FeCrAl alloy could be used for the cladding of commercial nuclear fuel. The use of this alloy will benefit the public since it is going to make the power generating light water reactors safer. In the Phase 1A of this cost shared project, GE (GRC + GNF) teamed with the University of Michigan, Los Alamos National Laboratory, Brookhaven National Laboratory, Idaho National Laboratory, and Oak Ridge National Laboratory to study the environmental and mechanical behavior of more than eight candidate cladding materials both under normal operation conditions of commercial nuclear reactors and under accident conditions in superheated steam (loss of coolant condition). The main findings are as follows: (1) Under normal operation conditions the candidate alloys (e.g. APMT, Alloy 33) showed excellent resistance to general corrosion, shadow corrosion and to environmentally assisted cracking. APMT also showed resistance to proton irradiation up to 5 dpa. (2) Under accident conditions the selected candidate materials showed several orders of magnitude improvement in the reaction with superheated steam as compared with the current zirconium based alloys. (3) Tube fabrication feasibility studies of FeCrAl alloys are underway. The aim is to obtain a wall thickness that is below 400 µm. (4) A strategy is outlined for the regulatory path approval and for the insertion of a lead fuel assembly in a commercial reactor by 2022. (5) The GE team worked closely with INL to have four rodlets tested in the ATR. GE provided the raw stock for the alloys, the fuel for the rodlets and the cost for fabrication/welding of the rodlets. INL fabricated the rodlets and the caps and welded them to

  15. Ferritic Alloys as Accident Tolerant Fuel Cladding Material for Light Water Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Rebak, Raul B. [General Electric Global Research, Schnectady, NY (United States)

    2014-09-30

    The objective of the GE project is to demonstrate that advanced steels such as iron-chromium-aluminum (FeCrAl) alloys could be used as accident tolerant fuel cladding material in commercial light water reactors. The GE project does not include fuel development. Current findings support the concept that a FeCrAl alloy could be used for the cladding of commercial nuclear fuel. The use of this alloy will benefit the public since it is going to make the power generating light water reactors safer. In the Phase 1A of this cost shared project, GE (GRC + GNF) teamed with the University of Michigan, Los Alamos National Laboratory, Brookhaven National Laboratory, Idaho National Laboratory, and Oak Ridge National Laboratory to study the environmental and mechanical behavior of more than eight candidate cladding materials both under normal operation conditions of commercial nuclear reactors and under accident conditions in superheated steam (loss of coolant condition). The main findings are as follows: (1) Under normal operation conditions the candidate alloys (e.g. APMT, Alloy 33) showed excellent resistance to general corrosion, shadow corrosion and to environmentally assisted cracking. APMT also showed resistance to proton irradiation up to 5 dpa. (2) Under accident conditions the selected candidate materials showed several orders of magnitude improvement in the reaction with superheated steam as compared with the current zirconium based alloys. (3) Tube fabrication feasibility studies of FeCrAl alloys are underway. The aim is to obtain a wall thickness that is below 400 µm. (4) A strategy is outlined for the regulatory path approval and for the insertion of a lead fuel assembly in a commercial reactor by 2022. (5) The GE team worked closely with INL to have four rodlets tested in the ATR. GE provided the raw stock for the alloys, the fuel for the rodlets and the cost for fabrication/welding of the rodlets. INL fabricated the rodlets and the caps and welded them to

  16. A mechanism for the formation of equiaxed grains in welds of aluminum-lithium alloy 2090

    International Nuclear Information System (INIS)

    Lin, D.C.; Wang, G.-X.; Srivatsan, T.S.

    2003-01-01

    In this technical note, the formation and presence of a zone of equiaxed grains (EQZ) along the fusion boundary of welded aluminum-lithium alloy 2090 using filler metals containing zirconium and lithium is presented and discussed. However, no EQZ was evident in welded joints of alloy 2090 using the commercial filler metals: aluminum alloy 2319 and 4145. Under identical conditions, aluminum-lithium alloy 2090 was fusion welded using several new filler metals containing various amounts of zirconium and lithium. Results reveal an increase in the width of the zone of equiaxed grains with an increase in zirconium and lithium content in the filler metal. A viable mechanism for the formation of equiaxed grains and its relationship to filler metal composition is highlighted

  17. Precipitation structures and mechanical properties of Al-Li-Zr alloy containing V

    International Nuclear Information System (INIS)

    Ying, J.K.; Ohashi, T.

    1999-01-01

    It is known that Al-Li alloys possess high elastic modulus and low density, and the metastable δ' (Al 3 Li) precipitate in these alloys affords considerable strengthening effect. However, with the strengthening resulting from the precipitation of δ' which is coherent with the matrix, these alloys suffer from low ductility and fracture toughness. It seems that the loss of ductility is the slip localization which occurs as a result of slip planes during deformation in connection with the specific hardening mechanism. As a result it indicates typical intergranular fracture. On the one hand, zirconium is used in many aluminum alloys to inhibit recrystallization during alloy processing. When zirconium is present in the alloy grain refinement occurs, which consequently, is considered as a factor that reduces the slip distance, and lowers the stress concentration across grain boundaries and at grain boundary triple points. Nevertheless, if only zirconium is added in Al-Li alloy it still shows intergranular fracture. By Zedaris et al., equilibrium phase Al 3 (Zr,V) in Al-Zr alloy containing V reduces the lattice mismatch along the c-axis with Al and, the L1 2 -structure metastable precipitates Al 3 (Zr,V) in Al-Zr-V alloys are stable at elevated temperature. Therefore, it is interesting to elucidate the effect of V in Al-Li-Zr alloy at the precipitation structures and mechanical properties of these alloys

  18. Influence of microstructure on the thermal creep behaviour of zirconium alloys: experimental analysis and implementation of homogenization approaches

    International Nuclear Information System (INIS)

    Brenner, R.

    2001-01-01

    Zirconium alloys widely used in the nuclear industry can present thermomechanical variability of their behavior (especially for thermal creep) as a function of their microstructure. To have a better control of the mechanical behavior of these alloys and also to take into account the possible evolution of their fabrication process (chemical composition, thermal treatments,... ), it is important to have a modeling tool which help us to describe the relationship between the microstructure and the macroscopic behavior. This study contributes to establish a predictive modelling, based on an experimental analysis coupled with a homogenization approach of the thermal creep behavior of Zr alloys. The experimental analysis of the crystallographic texture effect for Zircaloy-4 alloys shows how the strain rate and stress exponent of the different glide systems are anisotropic. Transmission Electronic Microscopy analysis have been undertaken in order to determine the link between the texture and the activated slip system considering various mechanical tests (Ioading paths). The experimental analysis for Zr-Nb-1%-O bring to evidence the solid solution effect of Nb on the hardening of this alloy and the weak effect of the precipitates distribution on thermal creep behavior. An elasto-viscoplastic micromechanical modelling has been developed taking into account the microstructure effects on the macroscopic behavior of Zr alloys. The 'quasi-elastic' approximate of the self consistent scheme based on the affine formulation is proposed and compared with others and earlier formulations. The accuracy of this formulation for our study is demonstrated, as well as the from the scale transition point of view and the simple numerical resolution. A good agreement is found for the description of thermal creep behavior of Zircaloy-4 and Zr-Nb-1%-O alloys. The analysis of the results at a local scale (especially slip system secondary activities) gives the current limit for the description of

  19. Resistivity changes of some amorphous alloys undergoing nanocrystallization

    Science.gov (United States)

    Barandiarán, J. M.; Fernández Barquín, L.; Sal, J. C. Gómez; Gorría, P.; Hernando, A.

    1993-10-01

    The electrical resistivity of amorphous alloys with compositions: Fe 73.5Nb 3Cu 1Si 13.5B 9, Fe 86Zr 7Cu 1B 6 and Co 80Nb 8B 12 has been studied in the temperature range from 300 to 1100 K, where crystallization occurs. The products of crystallization and the grain size have been studied by X-ray diffraction. In a first step, all the alloys crystallize with small grains of a few nanometers in diameter (nanocrystalline state), and the resistivity behavior at this process accounts for the difference between the amorphous and nanocrystalline phases. The nanocrystalline phases are: α-Fe-Si, α-Fe and fcc Co for the three compounds studied respectively. A second process, at which grain growth and precipitation of intermetallic compounds and borides takes place, has been found for all the alloys. The resistivity is sensitive, not only to the total transformed sample amount, but to the topological distribution of the crystalline phases, and therefore shows a more complex behavior than other well established techniques, as differential scanning calorimetry. This supplementary information given by the resistivity is also discussed.

  20. Experimental study of the electric resistivity in Heusler alloys

    International Nuclear Information System (INIS)

    Kunzler, J.V.

    1980-01-01

    Electrical resistivity measurements have been performed in the Cu 2 Mn (A1sub(1-x) Snsub(x)) Heusler alloys, where x = 0, 0.05, 0.10 and 0.15, in the temperature range from 4.2 to 800 0 K. Measurements have also been made on the Ni 2 MnX Heusler asloys, with X = In, Sn or Sb, in the range from 4.2 to 300 0 K. The experimental curves clearly show the importance of the ferromagnetic character for the alloys resistivity. The results obtained for the copper alloys, as well as for the Ni 2 MnSn alloy, are in agreement with an interpretation in terms of Bloch-Gruneisen and spin-disorder models, and fail to provide evidences of s-d scattering for the conduction electrons. This is not the case for the Ni 2 MnIn and Ni 2 MnSb alloys, in which the presence of (s-d) interband electronic scattering process, via phonon, was detected. Specially for the two last alloys specific heat and electronic photo-emissivity experiments are suggested. (Author) [pt

  1. Ferritic Alloys with Extreme Creep Resistance via Coherent Hierarchical Precipitates.

    Science.gov (United States)

    Song, Gian; Sun, Zhiqian; Li, Lin; Xu, Xiandong; Rawlings, Michael; Liebscher, Christian H; Clausen, Bjørn; Poplawsky, Jonathan; Leonard, Donovan N; Huang, Shenyan; Teng, Zhenke; Liu, Chain T; Asta, Mark D; Gao, Yanfei; Dunand, David C; Ghosh, Gautam; Chen, Mingwei; Fine, Morris E; Liaw, Peter K

    2015-11-09

    There have been numerous efforts to develop creep-resistant materials strengthened by incoherent particles at high temperatures and stresses in response to future energy needs for steam turbines in thermal-power plants. However, the microstructural instability of the incoherent-particle-strengthened ferritic steels limits their application to temperatures below 900 K. Here, we report a novel ferritic alloy with the excellent creep resistance enhanced by coherent hierarchical precipitates, using the integrated experimental (transmission-electron microscopy/scanning-transmission-electron microscopy, in-situ neutron diffraction, and atom-probe tomography) and theoretical (crystal-plasticity finite-element modeling) approaches. This alloy is strengthened by nano-scaled L21-Ni2TiAl (Heusler phase)-based precipitates, which themselves contain coherent nano-scaled B2 zones. These coherent hierarchical precipitates are uniformly distributed within the Fe matrix. Our hierarchical structure material exhibits the superior creep resistance at 973 K in terms of the minimal creep rate, which is four orders of magnitude lower than that of conventional ferritic steels. These results provide a new alloy-design strategy using the novel concept of hierarchical precipitates and the fundamental science for developing creep-resistant ferritic alloys. The present research will broaden the applications of ferritic alloys to higher temperatures.

  2. 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

  3. Corrosion behaviors of SMART materials in the ammonia atmosphere

    International Nuclear Information System (INIS)

    Baek, J. H.; Lee, M. H.; Choi, B. S.; Kim, J. P.; Jung, Y. H.; Lee, D. J.

    1999-01-01

    The corrosion characteristics of the zirconium-based alloy(Low-Sn Zircaloy-4) and titanium-based alloys(PT-7M and ASTM Gr.2), which would be used for fuel cladding tube and steam generator tube in the SMART, were investigated at 360 deg C, 400 deg C, 500 deg C, and 520 deg C in the ammonia atmosphere. In all test conditions, the resistance to uniform and nodular corrosion of zirconium-based alloy was inferior to that of titanium-based alloys. In the case of 360 deg C test, the corrosion rate of zirconium-based alloy decreased slightly with increasing the ammonia concentration, while that of titanium-based alloys increased. The test results above 400 deg C showed that the corrosion resistance of PT-7M alloys was superior to that of ASTM Gr.2 alloy and was not influenced from the variation of ammonia concentration

  4. Filler metal alloy for welding cast nickel aluminide alloys

    Science.gov (United States)

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

    1998-03-10

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

  5. Electrical resistivity of V-Cr-Ti alloys

    Energy Technology Data Exchange (ETDEWEB)

    Zinkle, S.J.; Gubbi, A.N.; Eatherly, W.S. [Oak Ridge National Lab., TN (United States)

    1997-04-01

    Room temperature electrical resistivity measurements have been performed on vanadium alloys containing 3-6%Cr and 3-6%Ti in order to evaluate the microstructural stability of these alloys. A nonlinear dependence on Cr and Ti concentration was observed, which suggests that either short range ordering or solute precipitation (perhaps in concert with interstitial solute clustering) has occurred in V-6Cr-6Ti.

  6. Dilatometric studies on uranium-zirconium-fissium alloy

    International Nuclear Information System (INIS)

    Banerjee, Aparna; Kulkarni, S.G.; Kulkarni, R.V.; Kaity, Santu

    2012-01-01

    The knowledge of thermophysical properties of U-Zr alloys are important for modelling fuel behaviour in nuclear reactor. Fissium is an alloy that approximates the equilibrium concentration of the metallic fission product elements left by metallurgical reprocessing. Coefficient of thermal expansion (CTE) data is needed to calculate stresses occurring in fuel and cladding with change in temperature. Coefficient of thermal expansion can be utilized to determine the change of alloy density as a function of temperature. In the present investigation, thermophysical properties like coefficient of thermal expansion and density were determined using dilatometer for U-20wt.%Zr-5wt.%Fs alloy prepared by arc melting process. The microstructural investigation was carried out using scanning electron microscope

  7. Final report on: Grain size determination in zirconium alloys (IAEA Research Contract No. 6025/Rb.)

    International Nuclear Information System (INIS)

    Martinez M, E.

    1991-12-01

    In spite of the amount of research developed the knowledge still is far from complete and in this basis the International Atomic Energy Agency, (IAEA), by means of the Working Group on Water Reactor Fuel Performance and Technology, initiated, in 1990 the Coordinated Research Programme named Grain Size Determination In Zirconium Alloys. Several countries were invited to participate and to contribute to the main objective of the programme, which can be state as: To develop a unified metallographic technique capable to show the microstructure of zircaloy in a reproducible and uniform manner. To fulfill the objective the following goals were established: A. To measure the grain size and perform an statistical treatment, in samples prepared specifically to show different amounts of cold work, recrystallization and grain growth. B. To compare the results obtained by the different laboratories involved in the programme. C. Finally, after the Ugine meeting, also the determination of the recrystallization and grain growth kinetics. (Author)

  8. Investigation on wear resistance and corrosion resistance of electron beam cladding co-alloy coating on Inconel617

    Science.gov (United States)

    Liu, Hailang; Zhang, Guopei; Huang, Yiping; Qi, Zhengwei; Wang, Bo; Yu, Zhibiao; Wang, Dezhi

    2018-04-01

    To improve surface properties of Inconel 617 alloy (referred to as 617 alloy), co-alloy coating metallurgically bonded to substrate was prepared on the surface of 617 alloy by electron beam cladding. The microstructure, phase composition, microhardness, tribological properties and corrosion resistance of the coatings were investigated. The XRD results of the coatings reinforced by co-alloy (Co800) revealed the presence of γ-Co, CoCx and Cr23C6 phase as matrix and new metastable phases of Cr2Ni3 and Co3Mo2Si. These hypoeutectic structures contain primary dendrites and interdendritic eutectics. The metallurgical bonding forms well between the cladding layer and the matrix of 617 alloy. In most studied conditions, the co-alloy coating displays a better hardness, tribological performance, i.e., lower coefficient of frictions and wear rates, corrosion resistance in 1 mol L‑1 HCl solution, than the 617 alloy.

  9. Corrosion resistance and electrochemical potentiokinetic reactivation testing of some iron-base hardfacing alloys

    International Nuclear Information System (INIS)

    Cockeram, B.V.

    1999-01-01

    Hardfacing alloys are weld deposited on a base material to provide a wear resistant surface. Commercially available iron-base hardfacing alloys are being evaluated for replacement of cobalt-base alloys to reduce nuclear plant activation levels. Corrosion testing was used to evaluate the corrosion resistance of several iron-base hardfacing alloys in highly oxygenated environments. The corrosion test results indicate that iron-base hardfacing alloys in the as-deposited condition have acceptable corrosion resistance when the chromium to carbon ratio is greater than 4. Tristelle 5183, with a high niobium (stabilizer) content, did not follow this trend due to precipitation of niobium-rich carbides instead of chromium-rich carbides. This result indicates that iron-base hardfacing alloys containing high stabilizer contents may possess good corrosion resistance with Cr:C < 4. NOREM 02, NOREM 01, and NoCo-M2 hardfacing alloys had acceptable corrosion resistance in the as-deposited and 885 C/4 hour heat treated condition, but rusting from sensitization was observed in the 621 C/6 hour heat treated condition. The feasibility of using an Electrochemical Potentiokinetic Reactivation (EPR) test method, such as used for stainless steel, to detect sensitization in iron-base hardfacing alloys was evaluated. A single loop-EPR method was found to provide a more consistent measurement of sensitization than a double loop-EPR method. The high carbon content that is needed for a wear resistant hardfacing alloy produces a high volume fraction of chromium-rich carbides that are attacked during EPR testing. This results in inherently lower sensitivity for detection of a sensitized iron-base hardfacing alloy than stainless steel using conventional EPR test methods

  10. Quantification of the distribution of hydrogen by nuclear microprobe at the Laboratory Pierre Sue in the width of zirconium alloy fuel clad of PWR reactors

    International Nuclear Information System (INIS)

    Raepsaet, C.; Bossis, Ph.; Hamon, D.; Bechade, J.L.; Brachet, J.C.

    2007-01-01

    Among the analysis techniques by ions beams, the micro ERDA (Elastic Detection Analysis) is an interesting technique which allows the quantitative distribution of the hydrogen in materials. In particular, this analysis has been used for hydride zirconium alloys, with the nuclear microprobe of the Laboratory Pierre Sue. This probe allows the characterization of radioactive materials. The technique principles are recalled and then two examples are provided to illustrate the fuel clad behavior in PWR reactors. (A.L.B.)

  11. Optical microscopic observation of texture anisotropy on zirconium and its alloys

    International Nuclear Information System (INIS)

    Chamagne, Louis.

    1978-01-01

    A study of the polarisation variation produced by hexagonal metals, especially zirconium and its alloys, when the sample is turned in a plane perpendicular to the light beam illuminating it revealed a link between the texture of the sample and the shape of the curves obtained by measuring the light it reflects during rotation. The first part of this work was carried out on oriented monocrystals. The angle for which the maximum appears is shown to be directly related to the angle between the crystallographic C axis and the normal to the measurement plane. It is therefore possible to define the position of the C axis in the crystal. The second part is the practical application to polycrystalline materials deformed by rolling. Though calculations on the shape of the curves are out of the question for the moment it is easy to compare shapes obtained under well-defined conditions. Examples: - metal treated in the β form and cooled at controlled speed; - sample laminated in 2 directions and having a similar isotropy in both; - influence of a 100% lamination, one of the two directions being taken for reference. These curves show that a manufacture can be followed and modified as requested by the customer. In addition the method requires no specialised technicians and the apparatus can be fitted to all microscopes possessing polarised light [fr

  12. Innovative approaches in the manufacture of zirconium alloy components for PHWRs

    International Nuclear Information System (INIS)

    Rao, M.N.; Srivastava, R.K.

    2005-01-01

    Selection of an appropriate route for the fabrication of Zirconium alloy fuel components has a direct bearing on the quality of finished product. Many sophisticated and intricate processes such as vacuum arc melting, extrusion, hot rolling and cold working processes - swaging, drawing and sheet rolling are employed. Many advances were made in eddy current and ultrasonic evaluation to meet the stringent quality control requirement and locate the micro flaws. Emphasis was laid on achieving high recoveries and manufacture the product at minimum cost. Several creative and innovative processes were adopted particularly in the fabrication of end caps and spacers. The spacers were produced through the wire route and subsequently parting them into tiny spacers, which is entirely different from the conventional route of fabricating the sheets followed by blanking and coining. This has improved the material recovery and the lead time has been reduced substantially. The end caps used for the closure of clad tubes have to meet the most stringent quality requirements to avoid micro-flaws. The manufacturing processes adopted have direct influence on the integrity of the finished product. Special defect standards were developed to identify and eliminate micro-flaws and thereby ensure consistent and repetitive quality product. The paper brings out the above innovative approaches made in fabrication and quality control techniques in the manufacture of fuel components for PHWR fuel bundles. (author)

  13. A new model for the in-reactor corrosion of zirconium alloys

    Energy Technology Data Exchange (ETDEWEB)

    Cox, B [University of Toronto, ON (Canada). Centre for Nuclear Engineering

    1997-02-01

    Previous models for the in-reactor corrosion of zirconium alloys have assumed that the mechanism is a completely solid-state diffusion process, determined by the growth and breakdown of the protective oxide film. In-reactor kinetics have been related to out-reactor kinetics with the oxide-metal interface temperature calculated from effects of heat flux. Recent experimental results have suggested that oxide dissolution and reprecipitation may be a major process leading to the formation of thick porous oxide films in-reactor. The model described here is based on the dissolution of primary recoil tracks in the oxide as the primary process distinguishing in-reactor from out-reactor corrosion. The consequences of such a model would be a very different microscopic morphology of in-reactor and out-reactor thick films, a significant irradiation effect on non-heat transfer surfaces, and a change in the kinetics of the overall process. This model should be equally applicable to PWR and BWR water chemistries because of the amphoteric nature of ZrO{sub 2}, and the effects of LiOH should operate by an essentially identical mechanism. A reciprocal rate equation should fit these processes and with additive terms seems capable of accommodating all water chemistry effects, except for discontinuous processes such as nodular corrosion. (author). 60 refs, 6 figs.

  14. Corrosion resistance of Ni-Cr-Mo alloys. Chemical composition and metallurgical condition's effects

    International Nuclear Information System (INIS)

    Zadorozne, N.S.; Rebak, Raul B.

    2009-01-01

    Ni-Cr-Mo alloys offer an outstanding corrosion resistance in a wide variety of highly-corrosive environments. This versatility is due to the excellent performance of nickel in hot alkaline solutions and the beneficial effect of chromium and molybdenum in oxidizing and reducing conditions, respectively. Alloy C-22 (22 % Cr-13 % Mo-3% W) is a well known versatile member of this family. Due to its excellent corrosion resistance in a wide variety of environments, Alloy C-22 has been selected for the fabrication of the corrosion-resistant outer shell of the high-level nuclear waste container. The increasing demand of the industry for corrosion resistant alloys with particular properties of corrosion and mechanical resistance has led to the development of new alloys. Alloy C-22HS (Ni-21 % Cr-17 % Mo) is a new high-strength corrosion resistant material recently developed and introduced into the market. This alloy provides a corrosion resistance comparable with that of other C-type alloys, and it can also be age hardened to effectively double its yield strength. HASTELLOY HYBRID-BC1 (Ni-22 % Mo-15 % Cr) is a new development intended for filling the gap between Ni-Mo and Ni-Cr-Mo alloys. This novel alloy is able to withstand HCl and H 2 SO 4 , even in the presence of dissolved oxygen and other oxidizing species. Its resistance to chloride-induced pitting corrosion, crevice corrosion and stress corrosion cracking is also remarkable. Thermal aging of Ni-Cr-Mo alloys leads to microstructure changes depending on the temperature range and exposure time at temperature. A Long Range Ordering (LRO) reaction can occur in the range of 350 C degrees to 600 C degrees, producing an ordered Ni 2 (Cr,Mo) phase. This ordering reaction does not seem to affect the corrosion resistance and produces only a slight loss in ductility. LRO transformation is homogeneous and has proven to be useful to fabricate the age-hard enable Alloy C22-HS. Tetrahedral Close Packed (TCP) phases, like μ, σ and

  15. Influence de l’irradiation et de la radiolyse sur la vitesse et les mécanismes de corrosion des alliages de zirconium

    OpenAIRE

    Verlet , Romain

    2015-01-01

    The nuclear fuel of pressurized water reactors (PWR) in the form of uranium oxide UO2 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 (UO2 pe...

  16. 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

  17. Magnesium and related low alloys

    International Nuclear Information System (INIS)

    Bernard, J.; Caillat, R.; Darras, R.

    1959-01-01

    In the first part the authors examine the comparative corrosion of commercial magnesium, of a magnesium-zirconium alloy (0,4 per cent ≤ Zr ≤ 0,7 per cent) of a ternary magnesium-zinc-zirconium alloy (0,8 per cent ≤ Zn ≤ 1,2 per cent) and of english 'Magnox type' alloys, in dry carbon dioxide-free air, in damp carbon dioxide-free air, and in dry and damp carbon dioxide, at temperatures from 300 to 600 deg. C. In the second part the structural stability of these materials is studied after annealings, of 10 to 1000 hours at 300 to 450 deg. C. Variations in grain after these heat treatments and mechanical stretching properties at room temperature are presented. Finally various creep rate and life time diagrams are given for these materials, for temperatures ranging from 300 to 450 deg. C. (author) [fr

  18. Dynamic behavior of zirconium alloy E110 under submicrosecond shock-wave loading

    Directory of Open Access Journals (Sweden)

    Kazakov D.N.

    2015-01-01

    Full Text Available Stress waves have been measured under shock wave loading of zirconium alloy E110 samples with the 0.5 – 8 mm thickness at normal and elevated temperatures. Duration of shock loading pulses varied from ∼0.05 up to 1μs with the amplitude varying from 3.4 up to 23 GPa. Free-surface velocity profiles have been registered using VISAR and PDV interferometers with nanosecond resolution. Attenuation of the elastic precursor has been measured to determine plastic strain rate behind the elastic precursor front. The plastic strain rate was observed to decrease with propagation from 106 s−1 at the 0.46-mm distance down to 2 ⋅ 104 s−1 at the 8-mm distance. Spall strength has been measured under normal and elevated temperatures. Spall strength versus strain rate relationships have been constructed in the 105 s−1 – 106s−1 range. Under shock compression higher than 10.6 GPa, the three-wave configuration of the shock wave has been registered and the polymorphous α → ω transition is considered to be the reason of this phenomenon. This work was supported by State Atomic Energy Corporation “Rosatom” within State Contract # H.4x.44.90.13.1111

  19. Dynamic behavior of zirconium alloy E110 under submicrosecond shock-wave loading

    Science.gov (United States)

    Kazakov, D. N.; Kozelkov, O. E.; Mayorova, A. S.; Malyugina, S. N.; Mokrushin, S. S.; Pavlenko, A. V.

    2015-09-01

    Stress waves have been measured under shock wave loading of zirconium alloy E110 samples with the 0.5 - 8 mm thickness at normal and elevated temperatures. Duration of shock loading pulses varied from ˜0.05 up to 1μs with the amplitude varying from 3.4 up to 23 GPa. Free-surface velocity profiles have been registered using VISAR and PDV interferometers with nanosecond resolution. Attenuation of the elastic precursor has been measured to determine plastic strain rate behind the elastic precursor front. The plastic strain rate was observed to decrease with propagation from 106 s-1 at the 0.46-mm distance down to 2 ṡ 104 s-1 at the 8-mm distance. Spall strength has been measured under normal and elevated temperatures. Spall strength versus strain rate relationships have been constructed in the 105 s-1 - 106s-1 range. Under shock compression higher than 10.6 GPa, the three-wave configuration of the shock wave has been registered and the polymorphous α → ω transition is considered to be the reason of this phenomenon. This work was supported by State Atomic Energy Corporation "Rosatom" within State Contract # H.4x.44.90.13.1111

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

    Science.gov (United States)

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

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

  1. Corrosion resistance and microstructure characterization of rare-earth-transition metal-aluminum-magnesium alloys

    International Nuclear Information System (INIS)

    Banczek, E.P.; Zarpelon, L.M.C.; Faria, R.N.; Costa, I.

    2009-01-01

    This paper reports the results of investigation carried out to evaluate the corrosion resistance and microstructure of some cast alloys represented by the general formula: La 0.7-x Pr x Mg 0.3 Al 0.3 Mn 0.4 Co 0.5 Ni 3.8 (x = 0, 0.1, 0.3, 0.5, and 0.7). Scanning electron microscopy (SEM) and electrochemical methods, specifically, polarization curves and electrochemical impedance spectroscopy (EIS), have been employed in this study. The effects of Pr substitution on the composition of the various phases in the alloys and their corrosion resistance have been studied. The electrochemical results showed that the alloy without Pr and the one with total La substitution showed the highest corrosion resistance among the studied alloys. The corrosion resistance of the alloys decreased when Pr was present in the lowest concentrations (0.1 and 0.3), but for higher Pr concentrations (0.5 and 0.7), the corrosion resistance increased. Corrosion occurred preferentially in a Mg-rich phase.

  2. Electrical resistivity of Al-Cu liquid binary alloy

    Science.gov (United States)

    Thakor, P. P.; Patel, J. J.; Sonvane, Y. A.; Jani, A. R.

    2013-06-01

    Present paper deals with the electrical resistivity (ρ) of liquid Al-Cu binary alloy. To describe electron-ion interaction we have used our parameter free model potential along with Faber-Ziman formulation combined with Ashcroft-Langreth (AL) partial structure factor. To see the influence of exchange and correlation effect, Hartree, Taylor and Sarkar et al local field correlation functions are used. From present results, it is seen that good agreements between present results and experimental data have been achieved. Lastly we conclude that our model potential successfully produces the data of electrical resistivity (ρ) of liquid Al-Cu binary alloy.

  3. A Novel Zr-1Nb Alloy and a New Look at Hydriding

    Energy Technology Data Exchange (ETDEWEB)

    Robert D. Mariani; James I. Cole; Assel Aitkaliyeva

    2013-09-01

    A novel Zr-1Nb has begun development based on a working model that takes into account the hydrogen permeabilities for zirconium and niobium metals. The beta-Nb secondary phase particles (SPPs) in Zr-1Nb are believed to promote more rapid hydrogen dynamics in the alloy in comparison to other zirconium alloys. Furthermore, some hydrogen release is expected at the lower temperatures corresponding to outages when the partial pressure of H2 in the coolant is less. These characteristics lessen the negative synergism between corrosion and hydriding that is otherwise observed in cladding alloys without niobium. In accord with the working model, development of nanoscale precursors was initiated to enhance the performance of existing Zr-1Nb alloys. Their characteristics and properties can be compared to oxide-dispersion strengthened alloys, and material additions have been proposed to zirconium-based LWR cladding to guard further against hydriding and to fix the size of the SPPs for microstructure stability enhancements. A preparative route is being investigated that does not require mechanical alloying, and 10 nanometer molybdenum particles have been prepared which are part of the nanoscale precursors. If successful, the approach has implications for long term dry storage of used fuel and for new routes to nanoferritic and ODS alloys.

  4. Alloying element's substitution in titanium alloy with improved oxidation resistance and enhanced magnetic properties

    Energy Technology Data Exchange (ETDEWEB)

    Yu, Ang-Yang, E-mail: ayyu@imr.ac.cn; Wei, Hua; Hu, Qing-Miao; Yang, Rui

    2017-01-15

    First-principles method is used to characterize segregation and magnetic properties of alloyed Ti/TiO{sub 2}interface. We calculate the segregation energy of the doped Ti/TiO{sub 2} interface to investigate alloying atom's distribution. The oxidation resistance of Ti/TiO{sub 2} interface is enhanced by elements Fe and Ni but reduced by element Co. Magnetism could be produced by alloying elements such as Co, Fe and Ni in the bulk of titanium and the surface of Ti at Ti/TiO{sub 2} interface. The presence of these alloying elements could transform the non-magnetic titanium alloys into magnetic systems. We have also calculated the temperature dependence of magnetic permeability for the doped and pure Ti/TiO{sub 2} interfaces. Alloying effects on the Curie temperature of the Ti/TiO{sub 2} interface have been elaborated. - Highlights: • We consider the segregation of alloying atoms on the Ti(101¯0)/TiO{sub 2}(100) interface. • Alloying the Ti//TiO{sub 2} interface with Fe and Ni has a great advantage of improving the oxidation resistance. • Fe, Co and Nican enhance the magnetic properties of the investigated system. • The variation of permeability with temperature has been presented.

  5. The effects of zirconium and beryllium on microstructure evolution, mechanical properties and corrosion behaviour of as-cast AZ63 alloy

    International Nuclear Information System (INIS)

    Jafari, Hassan; Amiryavari, Peyman

    2016-01-01

    Alloying elements are able to strongly modify the microstructure characteristics of Mg–Al–Zn alloys which dominate mechanical and corrosion properties of the alloys. In this research, the individual effects of Zr and Be additions on the microstructure, mechanical and corrosion properties of as-cast AZ63 alloy were explored. The results revealed that the addition of Zr leads to microstructure refinement in as-cast AZ63 alloy, resulting in improved tensile and hardness properties. 0.0001 and 0.001 wt% Be containing cast AZ63 alloy exhibited microstructure coarsening, while morphological alteration from sixford symmetrical to irregular shape grain was observed for the alloy containing 0.01 and 0.1 wt% Be. No specific Be compound was detected. In addition, mechanical properties of AZ63 alloy containing Zr was improved due to the microstructure modification, while Be containing alloy responded reverse behaviour. The corrosion resistance of AZ63 alloy was improved after the addition of Zr and Be due to the grain refinement and passivation effects, respectively. However, when the Zr content exceeds 0.5 wt%, the formation of Al 2 Zr affected the corrosion resistance. In other words, AZ63–0.5Zr alloy provided the lowest corrosion rate.

  6. Iron-niobium-aluminum alloy having high-temperature corrosion resistance

    Science.gov (United States)

    Hsu, Huey S.

    1988-04-14

    An alloy for use in high temperature sulfur and oxygen containing environments, having aluminum for oxygen resistance, niobium for sulfur resistance and the balance iron, is discussed. 4 figs., 2 tabs.

  7. Determination of tungsten in high-alloy steels and heat resisting alloys by isotope dilution-spark source mass spectrometry

    International Nuclear Information System (INIS)

    Saito, Morimasa; Yamada, Kei; Okochi, Haruno; Hirose, Fumio

    1983-01-01

    Tungsten in high-alloy steels and heat-resisting alloys was determined by isotope dilution method combined with spark source mass spectrometry by using 183 W enriched tungsten. The spike solution was prepared by fusing tungsten trioxide in sodium carbonate. A high-alloy steel sample was dissolved in the mixture of sulfuric acid and phosphoric acid together with the spike solution; a sample of heat resisting alloy was similarly dissolved in the mixture of hydrochloric acid, nitric acid, sulfuric acid, and phosphoric acid. The solution was evaporated to give dense white fumes. Tungsten was separated from the residue by a conventional cinchonine salt-precipitation method. The salt was ignited, and the residue was mixed with graphite powder and pressed into electrodes. The isotope 183 W and 184 W were measured. The method was applied to the determination of tungsten in JSS and NBS standard high-alloy steels and JAERI standard nickel- and NBS standard cobalt-base heat resisting alloys containing more than 0.05% tungsten. The results were obtained with satisfactory precision and accuracy. However, the results obtained for JSS standard high- speed steels containing molybdenum tended to be significantly lower than the certified values. (author)

  8. Electrophoretic deposition of hybrid coatings on aluminum alloy by combining 3-aminopropyltrimethoxysilan to silicon–zirconium sol solutions for corrosion protection

    Energy Technology Data Exchange (ETDEWEB)

    Yu, Mei; Xue, Bing; Liu, Jianhua, E-mail: yumei@buaa.edu.cn; Li, Songmei; Zhang, You

    2015-09-01

    Electrophoretic deposition (EPD) silicon–zirconium organic–inorganic hybrid coatings were applied on LC4 aluminum alloy for corrosion protection. 3-Glycidoxypropyl-trimethoxysilane (GTMS) and Zirconium (IV) n-propoxide (TPOZ) were used as precursors. 3-Aminopropyl-trimethoxysilane (APS) was added to enhance the corrosion protective performance of the coatings. Scanning electron microscopy (SEM), energy dispersive X-ray spectroscopy (EDS) and Fourier transform infrared spectroscopy (FTIR) were employed to characterize morphology, microstructure and component. The results show that the addition of APS leads to the enhanced migration and deposition of positively charged colloidal particles on the surface of metal substrate, which results in the thickness increasing of coatings. However, loading an excessive amount of APS gives a heterogeneous coating surface. The corrosion protective performance of coatings were measured by electrochemical impedance spectroscopy (EIS) and potentiodynamic polarization. The results indicate that the addition of APS improves corrosion protective performance of coatings. The optimal addition content of APS is about 15%. The 15% APS coating is uniform and dense, as well as has good corrosion protective performance. The impedance value (1.58 × 10{sup 5} Ω·cm{sup 2}, at the lowest frequency) of 15% APS coating is half order of magnitude higher than that of coating without APS, and 15% APS coating always keeps the best corrosion protective performance with prolonged immersion time. This kind of coating is identified with “double-structure” properties based on the analysis of EIS and potentiodynamic polarization. Furthermore, the equivalent circuit results indicate that the intermediate oxide layer plays a main role in corrosion protection. - Highlights: • Electrophoretic deposition hybrid coatings are prepared on LC4 aluminum alloy. • 3-Aminopropyl-trimethoxysilane (APS) enhances the corrosion protective performance. • The

  9. Electrophoretic deposition of hybrid coatings on aluminum alloy by combining 3-aminopropyltrimethoxysilan to silicon–zirconium sol solutions for corrosion protection

    International Nuclear Information System (INIS)

    Yu, Mei; Xue, Bing; Liu, Jianhua; Li, Songmei; Zhang, You

    2015-01-01

    Electrophoretic deposition (EPD) silicon–zirconium organic–inorganic hybrid coatings were applied on LC4 aluminum alloy for corrosion protection. 3-Glycidoxypropyl-trimethoxysilane (GTMS) and Zirconium (IV) n-propoxide (TPOZ) were used as precursors. 3-Aminopropyl-trimethoxysilane (APS) was added to enhance the corrosion protective performance of the coatings. Scanning electron microscopy (SEM), energy dispersive X-ray spectroscopy (EDS) and Fourier transform infrared spectroscopy (FTIR) were employed to characterize morphology, microstructure and component. The results show that the addition of APS leads to the enhanced migration and deposition of positively charged colloidal particles on the surface of metal substrate, which results in the thickness increasing of coatings. However, loading an excessive amount of APS gives a heterogeneous coating surface. The corrosion protective performance of coatings were measured by electrochemical impedance spectroscopy (EIS) and potentiodynamic polarization. The results indicate that the addition of APS improves corrosion protective performance of coatings. The optimal addition content of APS is about 15%. The 15% APS coating is uniform and dense, as well as has good corrosion protective performance. The impedance value (1.58 × 10 5 Ω·cm 2 , at the lowest frequency) of 15% APS coating is half order of magnitude higher than that of coating without APS, and 15% APS coating always keeps the best corrosion protective performance with prolonged immersion time. This kind of coating is identified with “double-structure” properties based on the analysis of EIS and potentiodynamic polarization. Furthermore, the equivalent circuit results indicate that the intermediate oxide layer plays a main role in corrosion protection. - Highlights: • Electrophoretic deposition hybrid coatings are prepared on LC4 aluminum alloy. • 3-Aminopropyl-trimethoxysilane (APS) enhances the corrosion protective performance. • The coating

  10. Formation of ω-phase in Zr-4 at.% Cr alloy

    International Nuclear Information System (INIS)

    Dobromyslov, A.V.; Kazantseva, N.V.

    1996-01-01

    The ω-phase has been discovered in zirconium-base alloys with the transition metals of Period 4 of the Periodic Table only in Zr-V, Zr-Cr, and Zr-Cu alloys. The first mention about the ω-phase formation in Zr-Cr alloys was given for Zr-4.5 at.%. However, there were no experimental data that confirmed this fact. W.M. Rumball and F.G. Elder presented the X-ray results on the ω-phase formation in Zr-3.9 at.%Cr, but at the present time there are no electron microscope studies of the structure of the ω-phase in this system. Investigations of the features of the ω-phase formation, morphology of the ω-phase and the mechanism of its formation in the different zirconium-base alloys are necessary to establish the common features of the formation of structures with the metastable phases. The task of the present work is to study the conditions and features of the ω-phase formation in the Zr-Cr alloys and the effect of the eutectoid decomposition on the formation of ω-phase. This article is part of the detailed investigations of the feature and condition of the ω-phase formation in zirconium-base alloys with the transition metals of the groups I and V to VIII of the Periodic Table

  11. Determination of concentration and particles size of second phase in pipes from zirconia alloys

    International Nuclear Information System (INIS)

    Gordaya, A.P.

    2003-01-01

    Computer photos processing microstructure of pipes from zirconium alloys (Zr+1%Nb, Zirlo), obtaining from different method zirconium on different enterprises are given. The results of estimation represented in given work

  12. On the resistivity of metal-tellurium alloys for low concentrations of tellurium

    International Nuclear Information System (INIS)

    Gorecki, J.

    1982-04-01

    The resistivity and thermoelectric power of metal-tellurium liquid alloys have been discussed for the case of small tellurium concentration. Nearly free electron model of conduction band has been used. The rapid increase of resistivity in transition metal-tellurium alloys has been predicted. (author)

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

    Science.gov (United States)

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

    2018-05-01

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

  14. Influence of protecting gel film on oxidation of zirconium alloys

    Czech Academy of Sciences Publication Activity Database

    Frank, H.; Weishauptová, Zuzana; Vrtílková, V.

    2007-01-01

    Roč. 360, č. 3 (2007), s. 282-292 ISSN 0022-3115 R&D Projects: GA ČR GA106/04/0043 Institutional research plan: CEZ:AV0Z30460519 Keywords : fuel cladding * corrosion * Zirconium oxide Subject RIV: JF - Nuclear Energetics Impact factor: 1.643, year: 2007

  15. Correlation between the resistivity and the atomic clusters in liquid Cu-Sn alloys

    Science.gov (United States)

    Jia, Peng; Zhang, Jinyang; Hu, Xun; Li, Cancan; Zhao, Degang; Teng, XinYing; Yang, Cheng

    2018-05-01

    The liquid structure of CuxSn100-x (x = 0, 10, 20, 33, 40, 50, 60, 75, 80 and 100) alloys with atom percentage were investigated with resistivity and viscosity methods. It can be found from the resistivity data that the liquid Cu75Sn25 and Cu80Sn20 alloys had a negative temperature coefficient of resistivity (TCR), and liquid Cu75Sn25 alloy had a minimum value of -9.24 μΩ cm K-1. While the rest of liquid Cu-Sn alloys had a positive TCR. The results indicated that the Cu75Sn25 atomic clusters existed in Cu-Sn alloys. In addition, the method of calculating the percentage of Cu75Sn25 atomic clusters was established on the basis of resistivity theory and the law of conservation of mass. The Cu75Sn25 alloy had a maximum volume of the atomic clusters and a highest activation energy. The results further proved the existence of Cu75Sn25 atomic clusters. Furthermore, the correlation between the liquid structure and the resistivity was established. These results provide a useful reference for the investigation of liquid structure via the sensitive physical properties to the liquid structure.

  16. Atomistic studies of cation transport in tetragonal ZrO2 during zirconium corrosion

    International Nuclear Information System (INIS)

    Bai, Xian-Ming; Zhang, Yongfeng; Tonks, Michael R.

    2015-01-01

    Zirconium alloys are the major fuel cladding materials in current reactors. The water-side corrosion is a significant degradation mechanism of these alloys. During corrosion, the transport of oxidizing species in zirconium dioxide (ZrO 2 ) determines the corrosion kinetics. Previously, it has been argued that the outward diffusion of cations is important for forming protective oxides. In this work, the migration of Zr defects in tetragonal ZrO 2 is studied with temperature accelerated dynamics and molecular dynamics simulations. The results show that Zr interstitials have anisotropic diffusion and migrate preferentially along the [001] or c direction in tetragonal ZrO 2 . The compressive stresses can increase the Zr interstitial migration barrier significantly. The migration of Zr interstitials at a grain boundary is much slower than in a bulk oxide. The implications of these atomistic simulation results in the Zr corrosion are discussed. (authors)

  17. Novel Concepts for Damage-Resistant Alloys in Next Generation Nuclear Power Systems

    Energy Technology Data Exchange (ETDEWEB)

    Stephen M. Bruemmer; Peter L. Andersen; Gary Was

    2002-12-27

    The discovery of a damage-resistant alloy based on Hf solute additions to a low-carbon 316SS is the highlight of the Phase II research. This damage resistance is supported by characterization of radiation-induced microstructures and microchemistries along with measurements of environmental cracking. The addition of Hf to a low-carbon 316SS reduced the detrimental impact of radiation by changing the distribution of Hf. Pt additions reduced the impact of radiation on grain boundary segregation but did not alter its effect on microstructural damage development or cracking. Because cracking susceptibility is associated with several material characteristics, separate effect experiments exploring strength effects using non-irradiated stainless steels were conducted. These crack growth tests suggest that irradiation strength by itself can promote environmental cracking. The second concept for developing damage resistant alloys is the use of metastable precipitates to stabilize the microstructure during irradiation. Three alloys have been tailored for evaluation of precipitate stability influences on damage evolution. The first alloy is a Ni-base alloy (alloy 718) that has been characterized at low neutron irradiation doses but has not been characterized at high irradiation doses. The other two alloys are Fe-base alloys (PH 17-7 and PH 17-4) that have similar precipitate structures as alloy 718 but is more practical in nuclear structures because of the lower Ni content and hence lesser transmutation to He.

  18. Enhancement of wear and corrosion resistance of beta titanium alloy by laser gas alloying with nitrogen

    Science.gov (United States)

    Chan, Chi-Wai; Lee, Seunghwan; Smith, Graham; Sarri, Gianluca; Ng, Chi-Ho; Sharba, Ahmed; Man, Hau-Chung

    2016-03-01

    The relatively high elastic modulus coupled with the presence of toxic vanadium (V) in Ti6Al4V alloy has long been a concern in orthopaedic applications. To solve the problem, a variety of non-toxic and low modulus beta-titanium (beta-Ti) alloys have been developed. Among the beta-Ti alloy family, the quaternary Ti-Nb-Zr-Ta (TNZT) alloys have received the highest attention as a promising replacement for Ti6Al4V due to their lower elastic modulus and outstanding long term stability against corrosion in biological environments. However, the inferior wear resistance of TNZT is still a problem that must be resolved before commercialising in the orthopaedic market. In this work, a newly developed laser surface treatment technique was employed to improve the surface properties of Ti-35.3Nb-7.3Zr-5.7Ta alloy. The surface structure and composition of the laser-treated TNZT surface were examined by grazing incidence X-ray diffraction (GI-XRD) and X-ray photoelectron spectroscopy (XPS). The wear and corrosion resistance were evaluated by pin-on-plate sliding test and anodic polarisation test in Hanks' solution. The experimental results were compared with the untreated (or base) TNZT material. The research findings showed that the laser surface treatment technique reported in this work can effectively improve the wear and corrosion resistance of TNZT.

  19. Stress corrosion cracking of zirconium and its alloys in halogenide solutions

    International Nuclear Information System (INIS)

    Farina, Silvia B.

    2001-01-01

    A doctoral thesis developed at the corrosion labs in CNEA a few years ago showed that zirconium and Zircaloy-4 were susceptible to stress corrosion cracking (SCC) in chloride aqueous solutions at potentials above the pitting potential. However, the nature of the phenomenon was not elucidated. On the other hand, references about the subject were scarce and contradictory. The development of new SCC models, in particular, the surface mobility SCC mechanism suggested a review of zirconium and Zircaloy-4 SCC in halogenide aqueous solutions. This mechanism predicts that zirconium should be susceptible to SCC not only in chloride solutions but also in bromide and iodide solutions due to the low melting point of the surface compounds formed by the interaction between the metal and the environment. The present work was aimed to determine the conditions under which SCC takes place and the mechanism operating during this process. For that purpose, the effect of electrochemical potential, strain rate and temperature on the SCC susceptibility of both, zirconium and Zircaloy-4 in chloride, bromide and iodide solutions was investigated. It was observed that those materials undergo stress corrosion cracking only at potentials higher than the breakdown potential. The crack velocity increased slightly with the applied potential, and the strain rate had an accelerating effect on the crack propagation rate. In both materials two steps were found during cracking. The first one was characterized as intergranular attack assisted by stress due to an anodic dissolution process. This step is followed by a transition to a transgranular mode of propagation, which was considered as the 'true' stress corrosion cracking step. The intergranular attack is the rate-determining step due to the fact that the transgranular propagation rate is higher than the intergranular propagation rate. Several stress corrosion cracking mechanisms were analyzed to explain the transgranular cracking. The predictions

  20. Optimum schedules of difficult-to-form heat-resistant alloys forging

    International Nuclear Information System (INIS)

    Majzengel'ter, V.A.; Shuvalov, A.A.; Perevozov, A.S.

    2000-01-01

    The process of manufacturing half finished discs for hydroturbine engines from heat resistant difficulty deformed nickel, iron-nickel and cobalt alloys (EI435, EI868, VZh145-ID, EK79-ID, EK152-ID, EI826, EP648-VI) is described. The recommendations on the modes of forging the single-phase nonaging and double phase aging alloys are developed. The conclusion is made, that the first compressions of ingots shoved be accomplished by small press runs. The subsequent compressions should constituted not less than 8% during one run. The total compression of the ingot during one heating should be within the concrete alloy properties. With the purpose of obtaining uniform fine-grain structure the ingot heating during the last manufacturing cycle should be accomplished within the range of 1100-1130 deg C for the majority of heat resistant alloys [ru

  1. Aluminium-nickel-iron alloys resistant to corrosion by water at high temperature. Their basic properties - their improvement; Les alliages aluminium-nickel-fer resistant a la corrosion par l'eau a haute temperature. Bases de leurs proprietes - leur amelioration

    Energy Technology Data Exchange (ETDEWEB)

    Coriou, H; Fournier, R; Grall, L; Hure, J [Commissariat a l' Energie Atomique, Saclay (France).Centre d' Etudes Nucleaires

    1959-07-01

    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) [French] Apres un rappel de l'evolution des etudes sur ces alliages, montrant l'etablissement de leurs proprietes fondamentales et, en particulier, structurales, on etudie: 1 - Les processus de penetration de la corrosion. On expose les resultats de l'etude micrographique de l'interface metal-oxyde sur une serie d'alliages traites dans l'eau et la vapeur entre 350 et 395 deg. C. On verifie l'hypothese de l'attaque par poches de pression gazeuse et on met en evidence un second processus de progression en profondeur par ilots de corrosion de type

  2. On the temperature dependence of the excess resistivity in dilute volatile alloys

    International Nuclear Information System (INIS)

    Uray, L.; Vicsek, T.

    1978-01-01

    In recrystallized wires of many important refractory alloys, an appreciable part of the temperature dependence of the measured excess resistivity is related to the radial distribution of the volatile solutes (extrinsic temperature dependence). Both the extrinsic and the intrinsic part of the temperature dependence of the excess resistivity have been determined for dilute WFe, WCo and WRe alloys, by measuring the resistance as a function of temperature and the thickness of layers removed by electrothinning. In this way the parameters of the evaporation profiles were also determined. In the surface region at low temperatures the length scale of the inhomogeneity is comparable to the mean-free path. Therefore, the observed extrinsic temperature dependence of the excess resistivity was calculated directly from the Boltzmann equation. The WCo alloy is a Kondo system, since its resistivity shows a minimum a 20 K. (author)

  3. Fracture properties and heat resistance of ceramics consisting of microspheres of stabilized zirconium dioxide

    International Nuclear Information System (INIS)

    Krasulin, Yu.L.; Barinov, S.M.; Ivanov, A.B.; Timofeev, V.N.; Grevtsev, S.N.; Ivanov, D.A.

    1980-01-01

    Determined were effective specific fracture work, critical coefficient of stress intensity in the upper point of the fracture, strength and heat resistance during heat changes (20-1300 deg C) of the material produced by sintering stabilized zirconium dioxide microspheres. Dependence of these characteristics on granulometric composition of microspheres was determined. It was ascertained that the additional introduction of large microspheres into the bulk of small microspheres increased the metal fracture work. Specific work of material fracture progress exceeded specific work of fracture motion initiation. High value of fracture work together with high strength permits to use the material formed of microspheres as structural ceramics

  4. New corrosion resistant alloys on the base of titanium and high-chromium steels

    International Nuclear Information System (INIS)

    Tomashov, N.D.; Chernova, G.P.

    1975-01-01

    It is shown that stability of titanium alloys, with α-structure (OT-4, AT3,AT6) and high-strength α+β or pure β-structure (BT-14; BT-15), in hydrochloric acid solutions may be significantly improved due to additional alloying by minor additions of Pd(0,2%) similar to pure titanium. Additions of 0,2% Pd also significantly improve acid resistance of alloys of the Fe-Cr system. The highest corrosion resistance has Fe,40%Cr,0,2%Pd alloy. This alloy is stable in 20-40%H 2 SO 4 and 1% HCl at 100 deg C

  5. Effect of preparation techniques on creep characteristics of the Zr-2. 5% Nb alloy at temperatures of 673 to 823 K

    Energy Technology Data Exchange (ETDEWEB)

    Pahutova, M; Kreici, J; Polesna, M [Ceskoslovenska Akademie Ved, Brno. Ustav Fyzikalni Metalurgie

    1976-01-01

    The effect of the initial raw material - zirconium sponge or zirconium iodide - on some creep and stres-strain properties was studied on Zr-2.5%Nb alloy by a stress-strain test at constant crosshead speed and by strain-rate sensitivity testing. Dependence of the creep characteristics on cooling conditions after solution treatment was examined. Alloy made from Zr-sponge was used for measurement of steady-state creep rate on time to fracture dependence and steady-state creep rate on time to fracture with respect to the angle between rolling direction of alloy sheets and tensile axis. Transmission electron microscopy was used for structure study of both alloys after different heat treatment. Higher creep strength of the alloy made from iodide zirconium (after respective heat treatment) than that of the alloy made from Zr-sponge is discussed. Oxygen content and its effect on structural changes during heat treatment seems to be responsible for higher creep strength of the first alloy. On the other hand the difference of respective creep strengths is not so significant as to justify production of Zr-2.5%Nb alloy and perhaps of future high-strength Zr alloys (for applications in structural components in reactors in the temperature range of 673 to 773 K) from iodide zirconium. Results of creep and stress-strain (short time) testing are briefly discussed.

  6. Age hardening and creep resistance of cast Al–Cu alloy modified by praseodymium

    International Nuclear Information System (INIS)

    Bai, Zhihao; Qiu, Feng; Wu, Xiaoxue; Liu, Yingying; Jiang, Qichuan

    2013-01-01

    The effects of praseodymium on age hardening behavior and creep resistance of cast Al–Cu alloy were investigated. The results indicated that praseodymium facilitated the formation of the θ′ precipitates during the age process and improved the hardness of the Al–Cu alloy. Besides, praseodymium resulted in the formation of the Al 11 Pr 3 phase in the grain boundaries and among the dendrites of the modified alloy. Because of the good thermal stability of Al 11 Pr 3 phase, it inhibits grain boundary migration and dislocation movement during the creep process, which contributes to the improvement in the creep resistance of the modified alloy at elevated temperatures. - Highlights: • Pr addition enhances the hardness and creep resistance of the Al–Cu alloy. • Pr addition facilitates the formation of the θ′ precipitates. • Pr addition results in the formation of the Al11Pr3 phase in the Al–Cu alloy

  7. Degradation of the Mechanical Properties of Zirconium-base alloys due to Interaction with Hydrogen

    International Nuclear Information System (INIS)

    Bertolino, Graciela

    2001-01-01

    Security aspects and the purpose to extend the nuclear power plants lifetime motivate the renovated interest on the influence of the environment and radiation on the mechanical properties of in-reactor materials.Zirconium based alloys are the family of alloys most extensively used in nuclear core components.A consequence of the interaction of the in-reactor environment with these alloys is the formation of brittle phase Zr hydride, a process that greatly affects the component integrity.In this work we present a experimental study of the hydrogen influence on the Z ry-4 mechanical properties at different temperatures.As a complement we also present results of a finite elements simulations of the fracture process.We performed standard metallurgical and mechanical characterization in commercial Z ry-4 samples to obtain their basic properties. Different hydrogen pickup techniques were applied to obtain H concentration of charged samples between 10 and 2000 ppm, homogeneous or mainly localized at the crack tip zone.To obtain the fracture toughness of the alloys specimens were tested using elastoplastic fracture mechanics techniques.Specifically we implement J-integral methodology with partial unloading compliance measurements.Tests were performed in a temperature range of 20 to 200 o C.The negative influence of the H content on material toughness probed to be important even at very small concentrations, with an effect that decreases when temperature increases.While there was observed no change in the fracture mechanism in homogeneous charged samples, specimens charged under a superimposed stress field fractured by brittle mode when were tested at 20 to 70 o C. SEM observations of the crack growth, the fracture surface morphology and precipitates content showed the influence of the precipitates on fracture at different H concentrations.At least three stages with different fracture behavior depending on H content were identified.Complementary to the experimental work we

  8. Experimental study and numerical modeling of the plastic behavior of zirconium alloys under and after irradiation

    International Nuclear Information System (INIS)

    Drouet, Julie

    2014-01-01

    Recrystallized zirconium alloys are widely used as constitutive material of cladding tubes in Pressurized Water Reactors. During their lifetime in reactor, these materials are submitted to irradiation, creating a large amount of defects and changing their mechanical behavior. Despite the broad knowledge of macroscopic modifications due to irradiation, microscopic mechanisms involved remain partially known and understood. This study aims at understanding this issue using two different means, experimental and numerical, to investigate interactions between moving dislocations and dislocation loops created by irradiation. The experimental approach is based on irradiating with Zr ions Zircaloy-4 samples. Then, these samples are strained in a transmission electron microscope (TEM). Mobile dislocations interacting with irradiation induced loops are observed, following different mechanisms. Loops can act as strong obstacles to moving dislocations, pinning their further glide and hardening the material. Therefore, this type of mechanism participates in irradiation hardening. Dislocations absorbing loops have also been observed, showing the ability of dislocations to clear up defects. This mechanism explains the formation of clear bands observed in the material after irradiation and mechanical testings. The numerical approach is based on Dislocation Dynamics (DD) simulations of mobile dislocations gliding in prismatic or basal planes of the hexagonal close packed lattice and loops, using NUMODIS. The results of this study are consistent with a recent study of interactions of dislocations in a prismatic plane and loops studied by molecular dynamics. The counterpart of this study with gliding dislocations in the basal plane, performed only using DD simulations, show interesting explanations of the observed clear band formation in basal and prismatic planes, with broader channels in basal planes. A situation observed during in situ TEM experiments has been simulated using DD

  9. Evaluation of the electrochemical behavior of U2.5Zr7.5Nb and U3Zr9Nb uranium alloys in relation to the pH and the solution aeration

    International Nuclear Information System (INIS)

    Mansur, Fabio Abud; Santos, Ana Maria Matildes dos; Ferraz, Wilmar Barbosa; Figueiredo, Celia de Araujo

    2011-01-01

    The Centro de Desenvolvimento da Tecnologia Nuclear (CDTN) is developing, in cooperation with the Centro Tecnologico da Marinha (CTMSP), the advanced nuclear plate type fuel for the second core of the land-based reactor prototype of the Laboratorio de Geracao Nucleo-Eletrica (LABGENE). Recent investigations have shown that the fuel made of uranium-based niobium and zirconium alloys reaches the best performance relative to other fuels, e.g. UO 2 . Niobium and Zirconium also increase the corrosion resistance and the mechanical strength of the uranium alloys. By means of electrochemical techniques the corrosion behavior of alloys U 2 . 5 Zr 7.5 Nb and U 3 Zr 9 Nb, developed at CDTN and heat treated in the temperature range of 200 deg C to 600 deg C, was assessed. The effect of the parameters pH and solution aeration was studied as well as the influence of zirconium and niobium alloying elements in the corrosion of uranium. The techniques used were open circuit potential, electrochemical impedance and potentiodynamic anodic polarization at room temperature. The tests were performed in a three-electrode electrochemical cell with Ag/AgCl (3M KCl) as the reference electrode and a platinum plate as the auxiliary electrode. The potentiodynamic polarization curves of uranium and its alloys in acidic solutions showed regions with anodic currents limited by a passive film. The presence of niobium and zirconium contributed for the formation of this film. The impedance data showed the presence of two semicircles in the Bode diagram, indicating the occurrence of two distinct electrochemical processes. The data were fitted to an equivalent circuit model in order to obtain parameters of the electrochemical processes and evaluate the effect of the studied variables. (author)

  10. The effect of preparation techniques on creep characteristics of the Zr-2.5% Nb alloy at temperatures of 673 to 823 K

    International Nuclear Information System (INIS)

    Pahutova, M.; Krejci, J.; Polesna, M.

    1976-01-01

    The effect of the initial raw material - zirconium sponge or zirconium iodide - on some creep and stres-strain properties was studied on Zr-2.5%Nb alloy by a stress-strain test at constant crosshead speed and by strain-rate sensitivity testing. Dependence of the creep characteristics on cooling conditions after solution treatment was examined. Alloy made from Zr-sponge was used for measurement of steady-state creep rate on time to fracture dependence and steady-state creep rate on time to fracture with respect to the angle between rolling direction of alloy sheets and tensile axis. Transmission electron microscopy was used for structure study of both alloys after different heat treatment. Higher creep strength of the alloy made from iodide zirconium (after respective heat treatment) than that of the alloy made from Zr-sponge is discussed. Oxygen content and its effect on structural changes during heat treatment seems to be responsible for higher creep strength of the first alloy. On the other hand the difference of respective creep strengths is not so significant as to justify production of Zr-2.5%Nb alloy and perhaps of future high-strength Zr alloys (for applications in structural components in reactors in the temperature range of 673 to 773 K) from iodide zirconium. Results of creep and stress-strain (short time) testing are briefly discussed. (author)

  11. Hydride precipitation, fracture and plasticity mechanisms in pure zirconium and Zircaloy-4 at temperatures typical for the postulated loss-of-coolant accident

    International Nuclear Information System (INIS)

    Pshenichnikov, Anton; Stuckert, Juri; Walter, Mario

    2016-01-01

    Highlights: • All δ-hydrides in Zr and Zircaloy-4 have basal or pyramidal types of habit planes. • Seven orientation relationships for δ-hydrides in Zr matrix were detected. • Decohesion fracture mechanism of hydrogenated Zr was investigated by fractography. - Abstract: The results of investigations of samples of zirconium and its alloy Zircaloy-4, hydrogenated at temperatures 900–1200 K (typical temperatures for loss-of-coolant accidents) are presented. The analyses, based on a range of complementary techniques (X-ray diffraction, scanning electron microscopy, electron backscatter diffraction) reveals the direct interrelation of internal structure transformation and hydride distribution with the degradation of mechanical properties. Formation of small-scale zirconium hydrides and their bulk distribution in zirconium and Zircaloy-4 were investigated. Fractographical analysis was performed on the ruptured samples tested in a tensile machine at room temperature. The already-known hydrogen embrittlement mechanisms based on hydride formation and hydrogen-enhanced decohesion and the applicability of them in the case of zirconium and its alloys is discussed.

  12. Creep resistance in a new alloy based on Fe3Al

    International Nuclear Information System (INIS)

    Morris, D.G.

    199