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Sample records for zirconium alloy characterised

  1. 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)

  2. 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.

  3. 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)

  4. 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

  5. Characterisation of neutron irradiation damage in zirconium alloys - a 'Round Robin' experiment

    International Nuclear Information System (INIS)

    Kelly, P.M.; Blake, R.G.; Jostsons, A.

    1977-01-01

    The nature of the damage structure in the neutron-irradiated zirconium specimens supplied as part of an international 'Round Robin' experiment has been studied using transmission electron microscopy. The damage structure consists entirely of a/3 dislocation loops and no evidence has been found for c component loops. Both vacancy and interstitial loops were found in specimens where inside/outside contrast analysis was possible. Quantitative measurements of loop size distributions and loop concentrations are reported. All specimens exhibited corduroy contrast to varying degress. (author)

  6. 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.

  7. 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

  8. 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)

  9. 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.

  10. 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

  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. 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.

  13. 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.

  14. 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

  15. 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)

  16. 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)

  17. 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.)

  18. 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.)

  19. 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.

  20. 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)

  1. 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

  2. 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.

  3. 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

  4. 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

  5. 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)

  6. 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.

  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. 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

  9. 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

  10. 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)

  11. 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.

  12. 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

  13. 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

  14. 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)

  15. 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)

  16. 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

  17. 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

  18. 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)

  19. 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

  20. 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

  1. 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

  2. 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.)

  3. 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

  4. 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.)

  5. 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.)

  6. 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)

  7. 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

  8. 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)

  9. 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

  10. 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)

  11. 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)

  12. 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

  13. 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.

  14. 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)

  15. 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.)

  16. 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.

  17. 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.)

  18. 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

  19. 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

  20. 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.

  1. 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)

  2. 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

  3. 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.

  4. 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.)

  5. 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

  6. 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

  7. 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

  8. 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.

  9. 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

  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. 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.

  12. 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)

  13. 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

  14. 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

  15. 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

  16. 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.

  17. 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)

  18. Working hardening modelization in zirconium alloys

    International Nuclear Information System (INIS)

    Sanchez, P.; Pochettino, Alberto A.

    1999-01-01

    Working hardening effects on mechanical properties and crystallographic textures formation in Zr-based alloys are studied. The hardening mechanisms for different grain deformations and topological conditions of simple crystal yield are considered. Results obtained show that the differences in the cold rolling textures (L and T textures) can be related with hardening microstructural parameters. (author)

  19. 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

  20. 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

  1. 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)

  2. 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.

  3. 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

  4. 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

  5. METMET fuel with Zirconium matrix alloys

    International Nuclear Information System (INIS)

    Savchenko, A.; Konovalov, I.; Totev, T.

    2008-01-01

    The novel type of WWER-1000 fuel has been designed at A.A. Bochvar Institute. Instead of WWER-1000 UO 2 pelletized fuel rod we apply dispersion type fuel element with uniformly distributed high uranium content granules of U9Mo, U5Nb5Zr, U3Si alloys metallurgically bonded between themselves and to cladding by a specially developed Zr-base matrix alloy. The fuel meat retains a controllable porosity to accommodate fuel swelling. The optimal volume ratios between the components are: 64% fuel, 18% matrix, 18% pores. Properties of novel materials as well as fuel compositions on their base have been investigated. Method of fuel elements fabrication by capillary impregnation has been developed. The primary advantages of novel fuel are high uranium content (more than 15% in comparison with the standard UO 2 pelletized fuel rod), low temperature of fuel ( * d/tU) and serviceability under transient conditions. The use of the novel fuel might lead to natural uranium saving and reduced amounts of spent fuel as well as to optimization of Nuclear Plant operation conditions and improvements of their operation reliability and safety. As a result the economic efficiency shall increase and the cost of electric power shall decrease. (authors)

  6. 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

  7. 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.)

  8. 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

  9. 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

  10. 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.

  11. 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

  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. 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.)

  14. 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

  15. 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)

  16. 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.

  17. 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.

  18. 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

  19. 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)

  20. 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

  1. 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.

  2. 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.

  3. 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.

  4. 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

  5. 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

  6. 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.

  7. 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.

  8. 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

  9. 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)

  10. 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)

  11. 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)

  12. 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

  13. 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)

  14. 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

  15. 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.

  16. 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.

  17. 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.

  18. 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

  19. 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...

  20. 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)

  1. 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)

  2. 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)

  3. 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

  4. 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.

  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.

    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

  6. 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.

  7. 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

  8. 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.

  9. 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.

  10. 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

  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. 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

  13. 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)

  14. 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)

  15. 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)

  16. 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

  17. 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

  18. Zirconium

    Science.gov (United States)

    Bedinger, G.M.

    2013-01-01

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

  19. 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

  20. 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

  1. 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

  2. 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

  3. 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

  4. Synthesis and structural characterisation using Rietveld and pair distribution function analysis of layered mixed titanium-zirconium phosphates

    International Nuclear Information System (INIS)

    Burnell, Victoria A.; Readman, Jennifer E.; Tang, Chiu C.; Parker, Julia E.; Thompson, Stephen P.; Hriljac, Joseph A.

    2010-01-01

    Crystalline metal (IV) phosphates with variable zirconium-to-titanium molar ratios of general formula (Ti 1-x Zr x )(HPO 4 ) 2 .H 2 O have been prepared by precipitation of soluble salts of the metals with phosphoric acid and heating the amorphous solids in 12 M H 3 PO 4 in an autoclave. The new materials are structurally characterised by Rietveld analysis of synchrotron X-ray powder diffraction data and pair distribution function (PDF) analysis of high energy synchrotron X-ray total scattering data. A broad range of zirconium-titanium phosphate solid solutions were formed showing isomorphous substitution of titanium by zirconium in the α-titanium phosphate lattice and vice versa for titanium substitution into the α-zirconium phosphate lattice. In both cases the solubility is partial with the coexistence of two substituted phases observed in samples with nominal compositions between the solubility limits. - Graphical abstract: Layered phosphates of general formula (Ti 1-x Zr x )(HPO 4 ).H 2 O have been prepared by the hydrothermal treatment of amorphous gels in phosphoric acid and characterised by Rietveld analysis of high resolution synchrotron X-ray powder diffraction data and pair distribution function analysis of high energy synchrotron X-ray total scattering data.

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

    Science.gov (United States)

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

    2006-06-01

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

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

    International Nuclear Information System (INIS)

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

    2006-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Riccardi, B. [Associazione Euratom-ENEA sulla Fusione, CR Frascati, P.B. 65 00044 Frascati, Roma (Italy)]. E-mail: riccardi@frascati.enea.it; Montanari, R. [Dipartimento di Ingegneria Meccanica, Universita di Roma, Tor Vergata, 00133 Roma (Italy); Casadei, M. [Centro Sviluppo Materiali, 00100 Roma (Italy); Costanza, G. [Dipartimento di Ingegneria Meccanica, Universita di Roma, Tor Vergata, 00133 Roma (Italy); Filacchioni, G. [ENEA CR Casaccia, I-00060 S. M. di Galeria, Roma (Italy); Moriani, A. [Associazione Euratom-ENEA sulla Fusione, CR Frascati, P.B. 65 00044 Frascati, Roma (Italy)

    2006-06-30

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

  8. 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

  9. 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

  10. 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

  11. 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)

  12. 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

  13. 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)

  14. 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.

  15. 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)

  16. 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

  17. 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

  18. 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.)

  19. 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.

  20. 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

  1. 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.)

  2. 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)

  3. 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

  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. 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

  6. 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.

  7. 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

  8. Corrosion protection of zirconium surface based on Heusler alloy

    Czech Academy of Sciences Publication Activity Database

    Horáková, Kateřina; Cichoň, Stanislav; Lančok, Ján; Kratochvílová, Irena; Fekete, Ladislav; Sajdl, P.; Krausová, A.; Macák, J.; Cháb, Vladimír

    2017-01-01

    Roč. 89, č. 4 (2017), s. 553-563 ISSN 0033-4545 R&D Projects: GA MŠk LO1409; GA ČR(CZ) GA16-03085S; GA ČR GJ17-19910Y; GA ČR(CZ) GA15-05095S Institutional support: RVO:68378271 ; RVO:67985858 Keywords : electrochemistry * silicon * spectroscopy * SSC-2016 * surface chemistry * wate * zirconium Subject RIV: JI - Composite Materials OBOR OECD: Composites (including laminates, reinforced plastics, cermets, combined natural and synthetic fibre fabrics Impact factor: 2.626, year: 2016

  9. 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.

  10. 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.)

  11. 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.

  12. 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

  13. 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

  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. 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

  16. 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

  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. 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.

  19. 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.

  20. 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)

  1. 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.

  2. 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

  3. 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.

  4. 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

  5. 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

  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. Comparison of DHC behaviour of two zirconium alloys

    International Nuclear Information System (INIS)

    Ponzoni, Lucio; Mieza, Ignacio; Heras, Evangelina De Las; Domizzi, Gladys

    2011-01-01

    Delayed hydride cracking (DHC) is an important cracking mechanism that may occur in Zr alloys during service in water-cooled reactors. Two conditions must be attained to initiate DHC: the stress intensity factor must be higher than a threshold value called K IH and hydrogen concentration must exceed a critical value. Currently the pressure tubes for CANDU reactor are fabricated from Zr-2.5Nb, but another Zr-alloy, Excel was evaluated demonstrating similar values of K IH but higher DHC velocity. In this paper the critical hydrogen concentration of Excel tube was evaluated and compared with that of Zr-2.5Nb. Due to higher hydrogen solubility limits in Excel, its critical concentration for DHC initiation is 10-40 wppm over that of Zr-2.5Nb in the range of 150 to 300 deg C. (author)

  8. Extractive photometric determination of zirconium in magnetic alloys

    International Nuclear Information System (INIS)

    Kutyrev, I.M.; Chernysheva, G.M.; Basargin, N.N.; Mikheev, N.I.

    1996-01-01

    A method for extractive photometric determination of Zr in magnetic alloys is presented. Extractive system - trioctylamine in toluene -H 2 SO 4 -Zr ensure selective and rapid (in single extraction) separation of Zr from Fe(3), Fe(2), Co, Ni, Cu, Al, Ti, Cr(3), Mn, Si, P, Nb, and Ta. The reliability of the method is confirmed in determination of Zr in the standerd sample SS 132c

  9. 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

  10. Phase characterisation in spark plasma sintered TiPt alloy

    CSIR Research Space (South Africa)

    Chikosha, S

    2011-12-01

    Full Text Available stream_source_info chikosha_2011.pdf.txt stream_content_type text/plain stream_size 4354 Content-Encoding UTF-8 stream_name chikosha_2011.pdf.txt Content-Type text/plain; charset=UTF-8 PHASE CHARACTERISATION IN SPARK... to form “necks”  Radiant Joule heat and pressure drives “neck” growth and material transfer © CSIR 2006 www.csir.co.za Page 6 Objective  Produce TiPt alloy compacts by Spark plasma sintering (SPS) of equiatomic...

  11. 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.

  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. 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. 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.)

  15. Growth and characterization of oxide layers on zirconium alloys

    International Nuclear Information System (INIS)

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

    1996-01-01

    In the range 265-435 C Zr-2.5Nb corrosion takes place in two stages, as opposed to the cyclic behaviour of Zry-4. The Zry-4 corrosion stages are described by a single equation, in terms of the dense oxide layer thickness that decreases sharply at each transition. Tetragonal zirconia is present in the oxide layers of both alloys. In Zry-4, its volume fraction decreases as the oxide grows; it is barely discernible in Zr-2.5Nb in films below 1 μm, to later increase up to the transition. In both alloys, compressive stresses are developed associated with the oxide growth. Their relaxation at the transition correlates with the transformation of ZrO 2 (t) to ZrO 2 (m) and with the decrease of the dense oxide layer. In Zr-2.5Nb, oxide ridges form on the β-Zr phase filaments, at the very onset of film growth. The cyclic behaviour associated with the periodical breakdown of the dense oxide layer is therefore blurred, although optical microscopy shows that the scale retains the multilayered structure typical of Zry-4. (orig.)

  16. 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

  17. 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

  18. 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

  19. 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)

  20. 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)

  1. 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

  2. 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.

  3. Comparison of delayed hydride cracking behavior of two zirconium alloys

    Energy Technology Data Exchange (ETDEWEB)

    Ponzoni, L.M.E. [CNEA – Centro Atómico Constituyentes, Hidrógeno en Materiales, Av. Gral. Paz 1499, San Martín (B1650KNA), Bs. As. (Argentina); Mieza, J.I. [CNEA – Centro Atómico Constituyentes, Hidrógeno en Materiales, Av. Gral. Paz 1499, San Martín (B1650KNA), Bs. As. (Argentina); Instituto Sabato, UNSAM–CNEA, Av. Gral. Paz 1499, San Martín (B1650KNA), Bs. As. (Argentina); De Las Heras, E. [CNEA – Centro Atómico Constituyentes, Hidrógeno en Materiales, Av. Gral. Paz 1499, San Martín (B1650KNA), Bs. As. (Argentina); Domizzi, G., E-mail: domizzi@cnea.gov.ar [CNEA – Centro Atómico Constituyentes, Hidrógeno en Materiales, Av. Gral. Paz 1499, San Martín (B1650KNA), Bs. As. (Argentina); Instituto Sabato, UNSAM–CNEA, Av. Gral. Paz 1499, San Martín (B1650KNA), Bs. As. (Argentina)

    2013-08-15

    Delayed hydride cracking (DHC) is an important failure mechanism that may occur in Zr alloys during service in water-cooled reactors. Two conditions must be attained to initiate DHC from a crack: the stress intensity factor must be higher than a threshold value called K{sub IH} and, hydrogen concentration must exceed a critical value. Currently the pressure tubes for CANDU reactor are fabricated from Zr–2.5Nb. In this paper the critical hydrogen concentration for DHC and the crack velocity of a developmental pressure tube, Excel, was evaluated and compared with that of Zr–2.5Nb. The DHC velocity values measured in Excel were higher than usually reported in Zr–2.5Nb. Due to the higher hydrogen solubility limits in Excel, its critical hydrogen concentration for DHC initiation is 10–50 wppm over that of Zr–2.5Nb in the range of 150–300 °C.

  4. Microstructure of amorphous and crystalline zirconium alloys rapiddly solidified

    International Nuclear Information System (INIS)

    Monteiro, W.A.; Bezerra, G.H

    1986-01-01

    In this work we report microstructural studies of rapidly solification of Zr-30% at Cu alloy. This composition was chosen because it is the Zr rich limit of glass formation range. The ribbons were prepared by melt spinning system (cooling rate is estimated in 10 6 K/s) and the average thickness of the microscopy were prepared by double jet electropolishing to investigate the microstructure of the ribbon. It was observed amorphos and crystalline regions. In the crystalline regions occured a radial growth morphology with stress contrats. The beginning of solidification is a polimorphous reaction and the shape of the micrograins is similar to spherulitic form. The average diameter of the grains are 0,5 μm or less. (Author) [pt

  5. 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.

  6. A mechanism for the hydrogen uptake process in zirconium alloys

    International Nuclear Information System (INIS)

    Cox, B.

    1999-01-01

    Hydrogen uptake data for thin Zircaloy-2 specimens in steam at 300-400 C have been analysed to show that there is a decrease in the rate of uptake with respect to the rate of oxidation when the terminal solid solubility (TSS) of hydrogen in the metal is exceeded. In order for TSS to be reached during pre-transition oxidation a very thin 0.125 mm Zircaloy sheet was used. The specimens had been pickled initially removing all Zr 2 (Fe/Ni) particles from the initial surfaces, yet the initial hydrogen uptake rates were still much higher than for Zircaloy-4 or a binary Zr/Fe alloy that did not contain phases that dissolve readily during pickling. Cathodic polarisation at room temperature in CuSO 4 solution showed that small cracks or pores formed the cathodic sites in pre-transition oxide films. Some were at pits resulting from the initial dissolution of the Zr 2 (Fe/Ni) phase; others were not; none were at the remaining intermetallics in the original surface. These small cracks are thought to provide the ingress routes for hydrogen. A microscopic steam starvation process at the bottoms of these small cracks or pores, leading to the accumulation of hydrogen adjacent to the oxide/metal interface, and causing breakdown of the passive oxide forming at the bottom of the flaw, is thought to provide the mechanism for the hydrogen uptake process during both pre-transition and post-transition oxidation. (orig.)

  7. 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.

  8. 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

  9. 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.

  10. 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)

  11. 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

  12. 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.

  13. Developments in delayed hydride cracking in zirconium alloys

    International Nuclear Information System (INIS)

    Puls, Manfred P.

    2008-01-01

    Delayed hydride cracking (DHC) is a process of diffusion assisted localized hydride embrittlement at flaws or regions of high stress. Models of DHC propagation and initiation have been developed that capture the essential elements of this phenomenon in terms of parameters describing processes occurring at the micro-scale. The models and their predictions of experimental results applied to Zr alloys are assessed. The propagation model allows rationalization of the effect of direction of approach to temperature and of the effect of the state and morphology of the beta phase in Zr-2.5Nb on DHC velocity. The K I dependence of the DHC velocity can only be approximately rationalized by the propagation models. This is thought to be because these models approximate the DHC velocity by a constant and shape-invariant rate of growth of the hydride at the flaw and have not incorporated a coupling between the applied stress field due to the flaw alone and the precipitated hydrides that would result in a variation of the shape and density of the hydrided region with K I . Separately, models have been developed for DHC initiation at cracks and blunt flaws. Expressions are obtained for the threshold stress intensity factor, K IH , for DHC initiation at a crack. A model for K IH has been used to rationalize the experimental result that DHC initiation is not possible above a certain temperature, even when hydrides can form at the crack tip. For blunt flaws with root radii in the μm range, and engineering process zone procedure has been derived to determine the initiation conditions requiring that both a critical stress and a critical flaw tip displacement must be achieved for hydride fracture. The engineering process zone procedure takes account of the dependence of DHC initiation on the flaw's root radius. Although all of the foregoing models are capable of describing the essential features of DHC, they are highly idealized and in need of further refinement. (author)

  14. 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)

  15. 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)

  16. 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)

  17. 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

  18. 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)

  19. 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

  20. 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)

  1. 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)

  2. 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

  3. 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)

  4. 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

  5. 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

  6. 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

  7. 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.

  8. 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.

  9. 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.)

  10. 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)

  11. 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

  12. 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

  13. 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)

  14. 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

  15. 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.

  16. 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.

  17. Stress-corrosion cracking characterisation of the advanced aerospace Al–Li 2099-T86 alloy

    Energy Technology Data Exchange (ETDEWEB)

    Goebel, J., E-mail: jannik.goebel@hzg.de; Ghidini, T.; Graham, A.J.

    2016-09-15

    New alloy developments driven by aircraft industry have identified aluminium lithium (Al–Li) alloys as potential candidates for substitution of incumbent high strength aluminium alloys used for manufacturing spacecraft and launchers. Whereas properties like specific stiffness, strength and toughness are proven as superior when compared to those of currently adopted Al alloys, the Stress Corrosion Cracking (SCC) characteristics are still an open aspect if advanced Al–Li alloys are considered for space structural applications. The present paper provides a comprehensive characterisation of the Al–Li 2099-T86 SCC performances.

  18. Stress-corrosion cracking characterisation of the advanced aerospace Al–Li 2099-T86 alloy

    International Nuclear Information System (INIS)

    Goebel, J.; Ghidini, T.; Graham, A.J.

    2016-01-01

    New alloy developments driven by aircraft industry have identified aluminium lithium (Al–Li) alloys as potential candidates for substitution of incumbent high strength aluminium alloys used for manufacturing spacecraft and launchers. Whereas properties like specific stiffness, strength and toughness are proven as superior when compared to those of currently adopted Al alloys, the Stress Corrosion Cracking (SCC) characteristics are still an open aspect if advanced Al–Li alloys are considered for space structural applications. The present paper provides a comprehensive characterisation of the Al–Li 2099-T86 SCC performances.

  19. Coupling between Experimental Measurements and Finite Element Calculations for identification of crystallographic constitutive law. Application to zirconium alloys; Methode de couplage entre experimentations et simulations numeriques en vue de l'identification de lois de comportement intracristallin. Application aux alliages de zirconium

    Energy Technology Data Exchange (ETDEWEB)

    Dexet, M

    2006-10-15

    This thesis presents a methodology for multi scale coupling between the morphology and texture of a microstructure as has been characterised experimentally, and the results of mechanical strain field analysis. This methodology is based on a coupling between experimental characterisation of the microstructure, ex-situ mechanical tests, local strain field measurements performed at the grain scale, and finite element simulations. Then, a definition of a cost function is proposed in order to optimise the parameters of the crystallographic constitutive law. This method is applied to the studies of zirconium alloys in order to improve the understanding of their mechanical behaviour in relation with their microstructures, which is a key requirement for their use in the nuclear industries. This work was funded by the joint research program SMIRN between EDF, CEA and CNRS. (author)

  20. Incorporation of Integral Fuel Burnable Absorbers Boron and Gadolinium into Zirconium-Alloy Fuel Clad Material

    International Nuclear Information System (INIS)

    Sridharan, K.; Renk, T.J.; Lahoda, E.J.; Corradini, M.L

    2004-01-01

    Long-lived fuels require the use of higher enrichments of 235U or other fissile materials. Such high levels of fissile material lead to excessive fuel activity at the beginning of life. To counteract this excessive activity, integral fuel burnable absorbers (IFBA) are added to some rods in the fuel assembly. The two commonly used IFBA elements are gadolinium, which is added as gadolinium-oxide to the UO2 powder, and boron, which is applied as a zirconium-diboride coating on the UO2 pellets using plasma spraying or chemical vapor deposition techniques. The incorporation of IFBA into the fuel has to be performed in a nuclear-regulated facility that is physically separated from the main plant. These operations tend to be very costly because of their small volume and can add from 20 to 30% to the manufacturing cost of the fuel. Other manufacturing issues that impact cost and performance are maintaining the correct levels of dosing, the reduction in fuel melting point due to gadolinium-oxide additions, and parasitic neutron absorption at fuel's end-of-life. The goal of the proposed research is to develop an alternative approach that involves incorporation of boron or gadolinium into the outer surface of the fuel cladding material rather than as an additive to the fuel pellets. This paradigm shift will allow for the introduction of the IFBA in a non-nuclear regulated environment and will obviate the necessity of additional handling and processing of the fuel pellets. This could represent significant cost savings and potentially lead to greater reproducibility and control of the burnable fuel in the early stages of the reactor operation. The surface alloying is being performed using the IBEST (Ion Beam Surface Treatment) process developed at Sandia National Laboratories. IBEST involves the delivery of energetic ion beam pulses onto the surface of a material, near-surface melting, and rapid solidification. The non-equilibrium nature of such processing allows f or surface

  1. 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

  2. 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

  3. 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

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

    International Nuclear Information System (INIS)

    Ribis, J.

    2007-12-01

    The fuel rod cladding, strongly affected by microstructural changes due to irradiation such as high density of dislocation loops, is strained by the end-of-life fuel rod internal pressure and the potential release of fission gases and helium during dry storage. Within the temperature range that is expected during dry interim storage, cladding undergoes long term creep under over-pressure. So, in order to have a predictive approach of the behavior of zirconium alloys cladding in dry storage conditions it is essential to take into account: initial dislocation loops, thermal annealing of loops and creep straining due to over pressure. Specific experiments and modelling for irradiated samples have been developed to improve our knowledge in that field. A Zr-1%Nb-O alloy was studied using fine microstructural investigations and mechanical testing. The observations conducted by transmission electron microscopy show that the high density of loops disappears during a heat treatment. The loop size becomes higher and higher while their density falls. The microhardness tests reveal that the fall of loop density leads to the softening of the irradiated material. During a creep test, both temperature and applied stress are responsible of the disappearance of loops. The loops could be swept by the activation of the basal slip system while the prism slip system is inhibited. Once deprived of loops, the creep properties of the irradiated materials are closed to the non irradiated state, a result whose consequence is a sudden acceleration of the creep rate. Finally, a micro-mechanical modeling based on microscopic deformation mechanisms taking into account experimental dislocation loop analyses and creep test, was used for a predictive approach by constructing a deformation mechanism map of the creep behavior of the irradiated material. (author)

  5. 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.)

  6. 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.

  7. 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.

  8. 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.

  9. 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

  10. 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

  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. 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

  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. A study on the fractures of iodine induced stress corrosion cracking of new zirconium alloys

    International Nuclear Information System (INIS)

    Peng Qian; Zhao Wenjin; Li Weijun; Tang Zhenghua; Heng Xuemei

    2005-10-01

    The morphology and chemical compositions of I-SCC fractures of new zirconium alloys were investigated by SEM and EDXA. The feature on fracture surface for I-SCC samples, such as corrosion products, the secondary cracking, intergranular cracking and pseudo-cleavage transgranular cracking, have been observed. And the fluting, the typical characteristic of I-SCC also has been found. Intergranular cracking is visible at crack initiation stage and transgranular cracking is distinguished at crack propagation stage. The corrosion products are mainly composed of Zr and O; and I can be detected on the local pseudocleavage zone. The most of grooves on the fractures of relieved-stress annealing samples are parallel with the roll plane. The intergranular cracking in relieved-stress annealing samples is not obvious. When the test temperature increases, the activity of iodine increases and the stress on crack tip is easier to be released, thus the corrosion products on fracture also increase and intergranular cracking is visible. The partial pressure of iodine influents the thickness of corrosion products, and intergranular cracking is easier to be found when iodine partial pressure is high enough. (authors)

  15. 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)

  16. 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)

  17. 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

  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. 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.

  20. 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

  1. 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

  2. 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

  3. 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)

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

    Energy Technology Data Exchange (ETDEWEB)

    Montgomery, Robert, E-mail: robert.montgomery@pnnl.gov [Pacific Northwest National Laboratory (United States); Tomé, Carlos, E-mail: tome@lanl.gov [Los Alamos National Laboratory (United States); Liu, Wenfeng, E-mail: wenfeng.liu@anatech.com [ANATECH Corporation (United States); Alankar, Alankar, E-mail: alankar.alankar@iitb.ac.in [Indian Institute of Technology Bombay (India); Subramanian, Gopinath, E-mail: gopinath.subramanian@usm.edu [University of Southern Mississippi (United States); Stanek, Christopher, E-mail: stanek@lanl.gov [Los Alamos National Laboratory (United States)

    2017-01-01

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

  5. In-pile creep test technique for zirconium alloys examination in BR-10 reactor channels

    International Nuclear Information System (INIS)

    Pevchikh, Yu.M.; Kruglov, A.S.; Troyanov, V.M.

    2002-01-01

    The irradiation enhanced creep phenomenon was discovered in stainless steels as a specific physical process accompanying high-intensity neutron flux irradiation in fast reactors. IPPE is also experienced in irradiation creep test activities, studying different types of materials under irradiation in BR-10 fast reactor. Series of in-channel type test facilities were constructed and tested in BR-10 reactor's 'dry' channels in order to carry out full-scale instrumented examination regarded to in-pile creep behaviour of different reactor materials. As a result, a specific test technique, named 'Tensometric method', has been developed and experimentally proved to be power enough in order to investigate irradiation creep of materials right in situ under neutron irradiation. The main peculiarity of test facility, which is constructed to apply the tensometric method, consists in absence of any special deformation-measurement cell at all. The in-pile creep strain measurement technique developed at IPPE is based on the non-direct measurement of specimen's deformation (either linear tensile strain or angular twisting one), which directly affects the loaded draws' tension parameters. Starting from 1993, in-pile creep experiments to investigate in-reactor creep behaviour of E110 and E635 zirconium alloys were carried out in BR-10. Experimental results and data collected during more than 20-year of BR-10 in-reactor creep test experience can be assumed as a strong evidence that the tensometric technique is a powerful instrument, which can give a chance to study different irradiation effects on reactor materials directly under irradiation. (author)

  6. Creep behavior under internal pressure of zirconium alloy cladding oxidized in steam at high temperature

    International Nuclear Information System (INIS)

    Chosson, Raphael

    2014-01-01

    During hypothetical Loss-Of-Coolant-Accident (LOCA) scenarios, zirconium alloy fuel cladding tubes creep under internal pressure and are oxidized on their outer surface at high temperature (HT). Claddings become stratified materials: zirconia and oxygen-stabilized α phase, called α(O), are formed on the outer surface of the cladding whereas the inner part remains in the β domain. The strengthening effect of oxidation on the cladding creep behavior under internal pressure has been highlighted at HT. In order to model this effect, the creep behavior of each layer had to be determined. This study focused on the characterization of the creep behavior of the α(O) phase at HT, through axial creep tests performed under vacuum on model materials, containing from 2 to 7 wt.% of oxygen and representative of the α(O) phase. For the first time, two creep flow regimes have been observed in this phase. Underlying physical mechanisms and relevant microstructural parameters have been discussed for each regime. The strengthening effect due to oxygen on the α(O) phase creep behavior at HT has been quantified and creep flow equations have been identified. A ductile to brittle transition criterion has been also suggested as a function of temperature and oxygen content. Relevance of the creep flow equations for each layer, identified in this study or from the literature, has been discussed. Then, a finite element model, describing the oxidized cladding as a stratified material, has been built. Based on this model, a fraction of the experimental strengthening during creep is predicted. (author) [fr

  7. 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

  8. 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

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

    Energy Technology Data Exchange (ETDEWEB)

    Glazoff, Michael Vasily [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2014-10-01

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

  10. 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

  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. 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

  13. 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

  14. 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

  15. A cell for the controllable thermal treatment and electrochemical characterisation of single crystal alloy electrodes

    DEFF Research Database (Denmark)

    Bondarenko, Alexander S.; Stephens, Ifan E.L.; Chorkendorff, Ib

    2012-01-01

    be performed in this cell. These include preparation and basic voltammetric characterisation of Cu/Pt(111) near-surface and surface alloys where monolayer amounts of Cu are located in the 1st and 2nd layers, respectively. The cell can also be useful for “electrochemical atomic layer epitaxy” to assemble...... multilayers using repetitive underpotential deposition....

  16. Mechanical Characterisation and Biomechanical and Biological Behaviours of Ti-Zr Binary-Alloy Dental Implants

    Directory of Open Access Journals (Sweden)

    Aritza Brizuela-Velasco

    2017-01-01

    Full Text Available The objective of the study is to characterise the mechanical properties of Ti-15Zr binary alloy dental implants and to describe their biomechanical behaviour as well as their osseointegration capacity compared with the conventional Ti-6Al-4V (TAV alloy implants. The mechanical properties of Ti-15Zr binary alloy were characterised using Roxolid© implants (Straumann, Basel, Switzerland via ultrasound. Their biomechanical behaviour was described via finite element analysis. Their osseointegration capacity was compared via an in vivo study performed on 12 adult rabbits. Young’s modulus of the Roxolid© implant was around 103 GPa, and the Poisson coefficient was around 0.33. There were no significant differences in terms of Von Mises stress values at the implant and bone level between both alloys. Regarding deformation, the highest value was observed for Ti-15Zr implant, and the lowest value was observed for the cortical bone surrounding TAV implant, with no deformation differences at the bone level between both alloys. Histological analysis of the implants inserted in rabbits demonstrated higher BIC percentage for Ti-15Zr implants at 3 and 6 weeks. Ti-15Zr alloy showed elastic properties and biomechanical behaviours similar to TAV alloy, although Ti-15Zr implant had a greater BIC percentage after 3 and 6 weeks of osseointegration.

  17. 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

  18. 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)

  19. 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)

  20. 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

  1. 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

  2. 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

  3. 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

  4. 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)

  5. Microstructural characterisation of friction stir welding joints of mild steel to Ni-based alloy 625

    Energy Technology Data Exchange (ETDEWEB)

    Rodriguez, J. [Brazilian Nanotechnology National Laboratory (LNNano), P.O. Box 6192, Campinas, SP (Brazil); University of Campinas (UNICAMP), Campinas, SP (Brazil); Ramirez, A.J., E-mail: ramirezlondono.1@osu.edu [Brazilian Nanotechnology National Laboratory (LNNano), P.O. Box 6192, Campinas, SP (Brazil); University of Campinas (UNICAMP), Campinas, SP (Brazil); Department of Materials Science and Engineering, The Ohio State University — OSU, Columbus, OH 43221 (United States)

    2015-12-15

    In this study, 6-mm-thick mild steel and Ni-based alloy 625 plates were friction stir welded using a tool rotational speed of 300 rpm and a travel speed of 100 mm·min{sup −1}. A microstructural characterisation of the dissimilar butt joint was performed using optical microscopy, scanning and transmission electron microscopy, and energy dispersive X-ray spectroscopy (XEDS). Six different weld zones were found. In the steel, the heat-affected zone (HAZ) was divided into three zones and was composed of ferrite, pearlite colonies with different morphologies, degenerated regions of pearlite and allotriomorphic and Widmanstätten ferrite. The stir zone (SZ) of the steel showed a coarse microstructure consisting of allotriomorphic and Widmanstätten ferrite, degenerate pearlite and MA constituents. In the Ni-based alloy 625, the thermo-mechanically affected zone (TMAZ) showed deformed grains and redistribution of precipitates. In the SZ, the high deformation and temperature produced a recrystallised microstructure, as well as fracture and redistribution of MC precipitates. The M{sub 23}C{sub 6} precipitates, present in the base material, were also redistributed in the stir zone of the Ni-based alloy. TMAZ in the steel and HAZ in the Ni-based alloy could not be identified. The main restorative mechanisms were discontinuous dynamic recrystallisation in the steel, and discontinuous and continuous dynamic recrystallisation in the Ni-based alloy. The interface region between the steel and the Ni-based alloy showed a fcc microstructure with NbC carbides and an average length of 2.0 μm. - Highlights: • Comprehensive microstructural characterisation of dissimilar joints of mild steel to Ni-based alloy • Friction stir welding of joints of mild steel to Ni-based alloy 625 produces sound welds. • The interface region showed deformed and recrystallised fcc grains with NbC carbides and a length of 2.0 μm.

  6. 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

  7. 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

  8. 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

  9. 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

  10. 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.

  11. 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.

  12. Characterisation and modelling of defect formation in direct-chill cast AZ80 alloy

    Energy Technology Data Exchange (ETDEWEB)

    Mackie, D.; Robson, J.D.; Withers, P.J. [School of Materials, University of Manchester, Manchester M13 9PL (United Kingdom); Turski, M. [Magnesium Elektron UK, Rake Lane, Manchester, M27 8BF (United Kingdom)

    2015-06-15

    Wrought magnesium alloys for demanding structural applications require high quality defect free cast feedstock. The aim of this study was to first identify and characterise typical defects in direct chill cast magnesium–aluminium–zinc (AZ) alloy billet and then use modelling to understand the origins of these defects so they can be prevented. Defects were first located using ultrasonic inspection and were then characterised using X-ray computed tomography (XCT) and serial sectioning, establishing the presence of oxide films and intermetallic particles Al{sub 8}Mn{sub 5} in all defects. A model was developed to predict the flow patterns and growth kinetics of the intermetallic phases during casting, which influence the formation of defects. Simulation of the growth of the intermetallic particles demonstrated that precipitation from the liquid occurs in the mould. The combination of the entrained oxide films and intermetallic particles recirculates in the liquid metal and continues to grow, until large enough to settle, which is predicted to occur at the centre of the mould where the flow is the slowest. Based on these predictions, strategies to reduce the susceptibility to defect formation are suggested. - Highlights: • Casting defects in magnesium direct chill casting have been imaged and characterised in 3-dimensions. • The occurrences of co-located clusters of particles and oxide films have been characterised and explained. • A coupled model has been developed to help interpret the observed trend for defects located towards the centre of billets.

  13. Characterisation and modelling of defect formation in direct-chill cast AZ80 alloy

    International Nuclear Information System (INIS)

    Mackie, D.; Robson, J.D.; Withers, P.J.; Turski, M.

    2015-01-01

    Wrought magnesium alloys for demanding structural applications require high quality defect free cast feedstock. The aim of this study was to first identify and characterise typical defects in direct chill cast magnesium–aluminium–zinc (AZ) alloy billet and then use modelling to understand the origins of these defects so they can be prevented. Defects were first located using ultrasonic inspection and were then characterised using X-ray computed tomography (XCT) and serial sectioning, establishing the presence of oxide films and intermetallic particles Al 8 Mn 5 in all defects. A model was developed to predict the flow patterns and growth kinetics of the intermetallic phases during casting, which influence the formation of defects. Simulation of the growth of the intermetallic particles demonstrated that precipitation from the liquid occurs in the mould. The combination of the entrained oxide films and intermetallic particles recirculates in the liquid metal and continues to grow, until large enough to settle, which is predicted to occur at the centre of the mould where the flow is the slowest. Based on these predictions, strategies to reduce the susceptibility to defect formation are suggested. - Highlights: • Casting defects in magnesium direct chill casting have been imaged and characterised in 3-dimensions. • The occurrences of co-located clusters of particles and oxide films have been characterised and explained. • A coupled model has been developed to help interpret the observed trend for defects located towards the centre of billets

  14. 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

  15. 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

  16. Characterising palladium-silver and palladium-nickel alloy membranes using SEM, XRD and PIXE

    International Nuclear Information System (INIS)

    Keuler, J.N.; Lorenzen, L.; Sanderson, R.D.; Prozesky, V.; Przybylowicz, W.J.

    1999-01-01

    Palladium alloy membranes were prepared by successive electroless plating steps on an alumina-zirconia support membrane. Palladium, silver and nickel were deposited in layers and then the metal films were heat treated for 5 h in a hydrogen atmosphere at 650 deg. C. The topography of the metal coatings and cross-sections of the films (before and after heating) were characterised using scanning electron microscopy (SEM). XRD was used to determine the crystal phase of the alloy coatings. Both SEM and XRD provide only surface information and therefore micro-PIXE was used to extract depth information of the alloy coating. Concentration profiles across the thickness of the films were constructed to determine penetration of the coating into the support membrane pores during electroless plating and to investigate diffusion of coated layers during the heating step

  17. Photoelectrochemical properties and band structure of oxide films on zirconium-transition metal alloys

    International Nuclear Information System (INIS)

    Takahashi, Kazuo; Uno, Masayoshi; Okui, Mihoko; Yamanaka, Shinsuke

    2006-01-01

    The microalloying effects of 4d and 5d transition metals, M (M: Nb, Mo, Ta, W) on the photoelectrochemical properties, the flat band potential (U fb ) and the band gap energy (E g ), for zirconium oxide films were investigated by photoelectrochemical measurements and band calculation. Button ingots of zirconium-5 mol% M (M: Nb, Mo, Ta, W) were made from high-purity metals (99.9% purity) by arc melting in a purified argon atmosphere. These plate specimens were sealed into silica tubes in vacuum, and then homogenized at 1273 K for 24 h. Subsequently, these specimens were oxidized up to 1173 K. The photocurrent of each specimen was evaluated at room temperature under the irradiation of Xe lamp (500 W) through grating monochrometer and cut-off filter. 0.1 M Na 2 SO 4 solution was used as the electrolyte. The value of the flat band potential was higher and the value of the band gap energy was smaller than that of pure zirconium oxide film in all sample. It was found from the calculation by CASTEP code that the decreases in band gap energy of these oxide films was due to formation of 4d or 5d orbital of transition metals

  18. 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

  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. Corrosion of zirconium alloys in nuclear reactors: A model for irradiation induced enhancement by local radiolysis in the porous oxide

    Energy Technology Data Exchange (ETDEWEB)

    Lemaignan, C; Salot, R [CEA/DRN/DTP, CENG-SECC, Grenoble (France)

    1997-02-01

    An analysis has been undertaken of the various cases of local enhancement of corrosion rate of zirconium alloys under irradiation. It is observed that in most cases a strong emission of energetic {beta}{sup -} is present leading to a local energy deposition rate higher than the core average. This suggests that the local transient radiolytic oxidizing species produced in the coolant by the {beta}{sup -} 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, and in front of Pt inserts or Cu-rich cruds. It explains also the irradiation corrosion enhancement of Cu-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 ionizing species like {alpha} from Ni-rich alloys and fission products in homogeneous reactors. Due to the changes induced by the irradiation intensity on the concentration of the radiolytic species, the coolant chemistry, that controls the boundary conditions for oxide growth, has to be analyzed with respect to the local value of the energy deposition rate. An analysis has been undertaken which shows that, in a porous media, the water is exposed to a higher intensity than bulk water. This leads to a higher concentration of oxidizing radiolytic species at the root of the cracks of the porous oxide, and increases the corrosion rate under irradiation. This mechanism, deduced from the explanation proposed for localized irradiation enhanced corrosion, can be extended to the whole reactor core, where the general enhancement of Zr alloys corrosion under irradiation could be attributed to the general radiolysis in the porous zirconia. (author). 18 refs, 3 figs, 3 tabs.

  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. Phase characterisation and mechanical behaviour of Fe–B modified Cu–Zn–Al shape memory alloys

    Directory of Open Access Journals (Sweden)

    Kenneth Kanayo Alaneme

    2017-04-01

    Full Text Available The microstructures, phase characteristics and mechanical behaviour of Cu–Zn–Al alloys modified with Fe, B, and Fe–B mixed micro-alloying additions has been investigated. Cu–Zn–Al alloys were produced by casting with and without the addition of the microelements (Fe, B and Fe–B. The alloys were subjected to a homogenisation – cold rolling – annealing treatment schedule, before the alloys were machined to specifications for tensile test, fracture toughness, and hardness measurement. Optical, scanning electron microscopy and X-ray diffraction analysis were utilised for microstructural and phase characterisation of the alloys. A distinct difference in grain morphology was observed in the alloys produced – the unmodified alloy had predominantly needle-like lath martensite structure with sharp grain edges while significantly larger transverse grain size and curve edged/near elliptical grain shape was observed for the modified Cu–Zn–Al alloys. Cu–Zn with fcc structure was the predominant phase identified in the alloys while Cu–Al with bcc structure was the secondary phase observed. The hardness of the unmodified Cu–Zn–Al alloy was higher than that of the modified alloys with reductions in hardness ranging between 32.4 and 51.5%. However, the tensile strength was significantly lower than that of the modified alloy grades (28.37–52.74% increase in tensile strength was achieved with the addition of micro-alloying elements. Similarly, the percent elongation and fracture toughness (10–23% increase of the modified alloy was higher than that of the unmodified alloy grade. The modified alloy compositions mostly exhibited fracture features indicative of a fibrous micro-mechanism to crack initiation and propagation, characterised by the prevalence of dimpled rupture.

  3. 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

  4. In situ oxidation of zirconium binary alloys by environmental SEM and analysis by AFM, FIB, and TEM

    International Nuclear Information System (INIS)

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

    2010-01-01

    Binary Zr-alloys containing 1%Fe and 1% Ni (large precipitates) and 1% Cr and 0.6% Nb (small precipitates), as well as a pure Zr sample were exposed in situ at 130 Pa water vapour pressure at 415 o C in an environmental SEM. The surface topography and composition of each sample was characterised before in situ experiments, during and after oxidation. After oxidation the surface was characterised by SEM and EDS, AFM and TEM combined with EDS. Focused ion beam was used to prepare cross sections of the metal-oxide interface and for the preparation of TEM thin foils. The oxidation behaviour of precipitates for these alloying elements can be characterised into two large families, those which show a rapid oxidation and those which induce a delayed oxidation in comparison with the Zr-matrix. At 415 o C after 1 h of oxidation for Zr1%Fe and Zr1%Ni, the formation of protrusions could be detected at the surface, being related to underlying SPP in the oxide. On Zr1%Cr and Zr0.6%Nb unoxidised SPPs were observed in the oxide, close to the metal-oxide interface. These SPPs were, however, oxidised close to the outer surface of the oxide. The surface roughness was increased for all materials after in situ oxidation, however, only for Zr1%Fe and Zr1%Ni protrusions appeared on the surface during oxidation. It was subsequently demonstrated that these latter correspond to the position of SPPs. For Zr1%Fe the surface roughness increased more than in the other materials and on these protrusions small iron oxide crystals have been observed at the surface. These observations confirm that Fe has a different behaviour compared to the other SPP forming elements, and it diffuses out to the free surface of the material. These alloying elements being the constituents of the commercial alloys (Fe and Cr for Zircaloy-4; Fe, Cr and Ni for Zircaloy-2 and Nb for all Nb-containing alloys), this study allows to separate their individual influence and can allow a subsequent comparison to the behaviour

  5. 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.

  6. 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.)

  7. 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)

  8. Analysis of zirconium alloys using inductively-coupled plasma emission spectrometry

    International Nuclear Information System (INIS)

    White, G.F.; Pickford, C.J.

    1982-06-01

    As part of an interlaboratory collaborative exercise, certain trace and minor elements have been determined in a proposed zircaloy reference material using inductively-coupled plasma emission spectrometry. A dissolution procedure involving hydrochloric and hydrofluoric acids was used for determination of Hf, Cr, Fe and Sn. Data have also been obtained for Ni, Cu and Mn. Use of a high resolution monochromator in a scanning mode was found necessary for measurement of the emission intensities in order to resolve the spectral lines of interest from the intense and complex emission from the zirconium matrix. (author)

  9. Characterisation

    Science.gov (United States)

    2007-03-01

    Characterisation. In Nanotechnology Aerospace Applications – 2006 (pp. 4-1 – 4-8). Educational Notes RTO-EN-AVT-129bis, Paper 4. Neuilly-sur-Seine, France: RTO...the Commercialisation Processes Concept IDEA Proof-of- Principle Trial Samples Engineering Verification Samples Design Verification Samples...SEIC Systems Engineering for commercialisation Design Houses, Engineering & R&D USERS & Integrators SE S U R Integrators Fabs & Wafer Processing Die

  10. 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

  11. Evaluation of a Ductility after High Temperature Oxidation with the Three-Point Bend Test in Zirconium Alloys

    International Nuclear Information System (INIS)

    Jung, Yang Il; Park, Sang Yoon; Park, Jeong Yong; Jeong, Yong Hwan

    2010-01-01

    In a light water reactor, the fuel cladding play an important role of preventing leakage of radioactive materials into the coolant, and thus the mechanical integrity of the cladding should be guaranteed under the conditions of normal and transient operation. In the case of a loss of coolant accident (LOCA), the cladding is subjected to a high temperature oxidation which is finally quenched because of an emergency coolant reflooding into the core. In this situation, the current LOCA criteria consist of five separate requirements: i) peak cladding temperature, ii) maximum cladding oxidation, iii) maximum hydrogen generation, iv) coolable geometry, and v) long-term cooling. The claddings lose their ductility due to the microstructural phase transformation from beta to martensite alpha-prime. and hydrogen up-take after LOCA. Since the reduction in ductility can induce embrittlement of claddings, post-quench ductility is one of the major concerns in transient operation circumstances. For the analysis, usually ring compression test are performed on ring samples cut from the tube to examine the oxidized cladding ductility. However, the test would not be applicable to the platelet samples which are general form of a specimen for developing alloys. As a high burn-up fuel cladding materials, Zircaloys are being replaced by modern zirconium alloys such as ZIRLO, and M5. Korea has also developed a new fuel cladding material HANA (high performance alloy for nuclear application) by the Korea Atomic Energy Research Institute. Because of the different composition of the newer claddings in comparison with the conventional Zircaloy-4, the high temperature oxidation behavior and the ductility after the oxidation would be different, and the properties should be evaluated how much the newer claddings were improved

  12. Highlighting micrographic structures of uranium-zirconium alloys with 6 per cent of weight of Zr

    International Nuclear Information System (INIS)

    Bouleau, Maurice

    1961-01-01

    In order to study the transformation kinetics of U-Zr alloys with a Zr content of 6 per cent in weight, the authors searched for a slow enough electrolytic polishing bath, and for an attack and examination method to highlight martensite structures produced by austempering and water tempering, and ultra-fine decomposition structures obtained by austempering. The authors explain the choice of a perchloric-butyl glycol polishing bath, of an examination under polarized light or normal light after appropriate attacks. These studies are reported for annealed alloys, and for processed alloys with martensite or ultra-fine decomposition structures [fr

  13. The melting-diffusion correlation in the plutonium-zirconium alloys

    International Nuclear Information System (INIS)

    Zanghi, J.-P.; Calais, Daniel.

    1975-01-01

    The activation volumes for self-diffusion of Pu in b.c.c. PuZr alloys (10 and 40at%Zr) have been determined, the validity of Nachtrieb's melting-diffusion correlation was checked. Indeed, in the Pu-40at%Zr alloy, which has a pressure temperature phase diagram whose liquidus has a positive slope, the activation volume is positive, whereas in pure epsilon Pu where the slope is negative, the activation volume is negative. A self-diffusion mechanism in PuZr alloys is proposed [fr

  14. Joining of yttria-tetragonal zirconia polycrystal with an aluminum-zirconium alloy

    International Nuclear Information System (INIS)

    Rathner, R.C.; Green, D.J.

    1990-01-01

    Specimens of yttria-tetragonal zirconia polycrystal (Y-TZP) have been joined with an Al-5.8 wt% Zr alloy at temperatures of 900 degrees C and above. The braze alloy contained large needlelike precipitates of the intermetallic phase Al 3 Sr. It is shown that these large precipitates can aid in strengthening of the joint, especially if they are close to the interface. With decreasing layer thickness, the strengths increased with values as high as 420 MPa

  15. Twinning during β → α slow cooling in a zirconium alloy

    International Nuclear Information System (INIS)

    Luan, B.F.; Chai, L.J.; Wu, G.L.; Yu, H.B.; Chen, J.W.; Liu, Q.

    2012-01-01

    Twinning is frequently reported in Zr alloys during deformation and martensitic phase transformation. In the present work, however, we report {101 ¯ 2} and {112 ¯ 1} twinning after furnace cooling form β phase domain to room temperature in a Zr alloy. Intergranular thermal residual stresses produced during cooling due to anisotropic coefficients of thermal expansion, combined with the extraordinarily large grain size of the transformed α-Zr, are believed to be the reasons accounting for the formation of these twins.

  16. High Temperature Oxidation Behavior of Zirconium Alloy with Nano structured Oxide Layer in Air Environment

    International Nuclear Information System (INIS)

    Park, Y. J.; Kim, J. W.; Park, J. W.; Cho, S. O.

    2016-01-01

    If the temperature of the cladding materials increases above 1000 .deg. C, which can be caused by a loss of coolant accident (LOCA), Zr becomes an auto-oxidation catalyst and hence produces a huge amount of hydrogen gas from water. Therefore, many investigations are being carried out to prevent (or reduce) the hydrogen production from Zr-based cladding materials in the nuclear reactors. Our team has developed an anodization technique by which nanostructured oxide can be formed on various flat metallic elements such as Al, Ti, and Zr-based alloy. Anodization is a simple electrochemical technique and requires only a power supply and an electrolyte. In this study, Zr-based alloys with nanostructured oxide layers were oxidized by using Thermogravimetry analysis (TGA) and compared with the pristine one. It reveals that the nanostructured oxide layer can prevent oxidation of substrate metal in air. Oxidation behavior of the pristine Zr-Nb-Sn alloy and the Zr-Nb-Sn alloy with nanostructured oxide layer evaluated by measuring weight gain (TGA). In comparison with the pristine Zr-Nb-Sn alloy, weight gain of the Zr-Nb-Sn alloy with nanostructured oxide layer is lower than 10% even for 12 hours oxidation in air.

  17. Microstructural characterisation of Al-Si cast alloys containing rare earth additions

    Science.gov (United States)

    Elgallad, E. M.; Ibrahim, M. F.; Doty, H. W.; Samuel, F. H.

    2018-05-01

    This paper presents a thorough study on the effect of rare earth elements, specifically La and Ce, on the microstructure characteristics of non-modified and Sr-modified A356 and A413 alloys. Several alloys were prepared by adding 1% La and 1% Ce either individually or in combination. Microstructural characterisation was carried out using optical microscopy, scanning electron microscopy and electron probe microanalysis as well as differential scanning calorimetry (DSC) analysis. The results showed that the individual and combined additions of La and Ce did not bring about any modification or even refinement in the eutectic Si structure. Moreover, these additions were found to negate the modification effect of Sr, particularly in the presence of La. The A356 and A413 alloys containing La and/or Ce displayed high phase volume fractions owing to the formation of Al-Si-La/Ce/(La,Ce) and Al-Ti-La/Ce intermetallic phases. DSC analysis revealed that the formation temperatures of these phases varied from 560 to 568 °C and 568 to 574 °C, respectively. This analysis also showed that the addition of La and Ce whether individually or in combination resulted in a depression in the eutectic temperature and a considerable increase in the solidification range, particularly for the A413 alloy.

  18. 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.

  19. 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

  20. Tungsten-zirconium carbide-rhenium alloys with extraordinary thermal stability

    Energy Technology Data Exchange (ETDEWEB)

    Yang, X.D. [Key Laboratory of Materials Physics, Institute of Solid State Physics, Chinese Academy of Sciences, Hefei 230031 (China); Xie, Z.M.; Miao, S. [Key Laboratory of Materials Physics, Institute of Solid State Physics, Chinese Academy of Sciences, Hefei 230031 (China); University of Science and Technology of China, Hefei 230026 (China); Liu, R.; Jiang, W.B. [Key Laboratory of Materials Physics, Institute of Solid State Physics, Chinese Academy of Sciences, Hefei 230031 (China); Zhang, T., E-mail: zhangtao@issp.ac.cn [Key Laboratory of Materials Physics, Institute of Solid State Physics, Chinese Academy of Sciences, Hefei 230031 (China); Wang, X.P., E-mail: xpwang@issp.ac.cn [Key Laboratory of Materials Physics, Institute of Solid State Physics, Chinese Academy of Sciences, Hefei 230031 (China); Fang, Q.F. [Key Laboratory of Materials Physics, Institute of Solid State Physics, Chinese Academy of Sciences, Hefei 230031 (China); University of Science and Technology of China, Hefei 230026 (China); Liu, C.S., E-mail: csliu@issp.ac.cn [Key Laboratory of Materials Physics, Institute of Solid State Physics, Chinese Academy of Sciences, Hefei 230031 (China); Luo, G.N. [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); Liu, X. [Southwest Institute of Plasma Physics, Chengdu (China)

    2016-05-15

    The low recrystallization temperature (1200 °C) of pure W is a serious limitation for application as facing plasma materials in fusion reactor. In this paper, W-0.5wt.%ZrC-1wt.%Re (WZR) alloy with recrystallization temperature up to 1800 °C was prepared by mechanical milling and spark plasma sintering. The grain size of WZR alloy is about 2.6 μm, smaller than that of pure W (4.4 μm), which keeps unchanged until the annealing temperature increases to 1800 °C. Tensile tests indicate that the WZR alloys exhibit excellent comprehensive properties: the ductile to brittle transition temperature of WZR is in the range from 400 °C to 500 °C, about 200 °C lower than that of pure W prepared by the same process; the total elongation (TE) of WZR at 600 °C is above 30%, which is about 2 times that of pure W (at 700 °C). Meanwhile its tensile strength keeps ∼450 MPa before and after 1800 °C annealing as well as its TE increases after annealing. WZR alloy exhibits higher hardness (489HV) than that of pure W (453HV) at room temperature. Microstructure analysis indicates that the strengthening of nano-sized ZrC particles dispersion and Re solid solution improve tensile properties and thermal stability of WZR alloy.

  1. Zirconium hydrides and Fe redistribution in Zr-2.5%Nb alloy under ion irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Idrees, Y.; Yao, Z. [Department of Mechanical and Materials Engineering, Queen' s University, Kingston, ON, Canada, K7L 3N6 (Canada); Cui, J.; Shek, G.K. [Kinetrics, Mississauga, ON (Canada); Daymond, M.R., E-mail: daymond@queensu.ca [Department of Mechanical and Materials Engineering, Queen' s University, Kingston, ON, Canada, K7L 3N6 (Canada)

    2016-11-15

    Zr-2.5%Nb alloy is used to fabricate the pressure tubes of the CANDU reactor. The pressure tube is the primary pressure boundary for coolant in the CANDU design and is susceptible to delayed hydride cracking, reduction in fracture toughness upon hydride precipitation and potentially hydride blister formation. The morphology and nature of hydrides in Zr-2.5%Nb with 100 wppm hydrogen has been investigated using transmission electron microscopy. The effect of hydrides on heavy ion irradiation induced decomposition of the β phase has been reported. STEM-EDX mapping was employed to investigate the distribution of alloying elements. The results show that hydrides are present in the form of stacks of different sizes, with length scales from nano- to micro-meters. Heavy ion irradiation experiments at 250 °C on as-received and hydrided Zr-2.5%Nb alloy, show interesting effects of hydrogen on the irradiation induced redistribution of Fe. It was found that Fe is widely redistributed from the β phase into the α phase in the as-received material, however, the loss of Fe from the β phase and subsequent precipitation is retarded in the hydrided material. This preliminary work will further the current understanding of microstructural evolution of Zr based alloys in the presence of hydrogen. - Graphical abstract: STEM HAADF micrographs at low magnification showing the hydride structure in Zr-2.5Nb alloy.

  2. Study on Recrystallization of Cold-worked and β-quenched zirconium alloys

    International Nuclear Information System (INIS)

    Goo, J. S.; Hong, S. I.; Kim, H. S.; Jeong, Y. H.

    1998-01-01

    The observation of microstructure and the hardness test of Zr-Sn binary and Zircaloy-4 alloys were performed to investigate the recrystallization of cold-worked and β-quenched Zr alloys. All specimens were heat-treated in vacuum condition at various temperatures. From the observation of microstructures of cold-worked and β-quenched Zr alloys, the cold-worked specimens were shown to keep the cold-worked micro- structure as annealing temperature increased up to 500 deg C and the recrystallization was completed at between 550 deg C and 700 deg C. Meanwhile, the recrystallization of β-quenched Zr alloys was started at about 700 deg C. In all specimens of cold-worked and β-quenched Zr alloys, the hardness value tended to be consistent with microstructure. Although the cold-worked and the β-quenched specimens had an equal initial hardness value, the recrystallization behavior was indicated to be different from each other, which means that recrystallization mechanism is different from each other

  3. 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.)

  4. INR participation in the IAEA research project investigating the influence of hydrogen absorption on zirconium alloy behavior

    International Nuclear Information System (INIS)

    Roth, Maria; Radu, Vasile; Dobrea, Dumitru; Pitigoi, Vasile

    2003-01-01

    The paper summarizes the results obtained at INR Pitesti from its participation in the research project coordinated by IAEA Vienna in cooperation with Chalk River and AECL Canada, titled 'Hydrogen and Hydride Induced Degradation of the Mechanical and Physical Properties of Zirconium-based Alloys'. Evidenced is the contribution of INR Pitesti in the works of this project as well as the benefits of this participation for Romania as owner of CANDU type reactor. In the frame this project new results concerning the propagation rate of DHC type cracks in pressure tubes in CANDU reactors were obtained. The same method used to investigate the DHC project was adapted for determination of other quantities of interest related to structural integrity of the materials. The methodology was applied for testing the pressure tubes in Cernavoda NPP Unit 1. The contribution of INR team to statistical processing of data obtained in all the laboratories participating in this project is also highlighted. Opportunity afforded by IAEA to INR Pitesti to bring its contribution to the development of this project of international cooperation together with other well-known institutions and the support from RAAN are acknowledged. These opened ways for other fruitful international cooperation

  5. 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

  6. 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

  7. 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

  8. Evidence of zirconium nano-agglomeration in as-cast dilute U–Zr alloys

    International Nuclear Information System (INIS)

    Mukherjee, S.; Kaity, S.; Saify, M.T.; Jha, S.K.; Pujari, P.K.

    2014-01-01

    Microstructure evaluation of as-cast and annealed U–Zr (Zr = 2, 6 and 10 wt.%) alloys has been carried out for the first time using positrons as a probe. The chemical signature in the matter–antimatter annihilation gamma and the positron lifetime data suggests that majority of positrons are annihilating from Zr sites in the as-cast alloys. The results have been interpreted as due to the presence of Zr nano-agglomerates in the as-cast alloys which have a higher positron affinity as compared to the rest of the U matrix. A minimum agglomerate size of ∼2 nm diameter has been calculated from the difference in positron affinity between the agglomerates and the matrix. Upon annealing, the Zr signature in the annihilation gamma photons vanishes suggesting that the Zr agglomerates diffuse out of U matrix and form micron-sized precipitates. This has been confirmed by scanning electron microscopy which shows a 3 times increase in the surface density of the precipitates in the annealed alloys as compared to the as-cast ones. Shorter positron diffusion length (measured using slow positron beam) as compared to precipitate separation has been invoked to explain the observed data

  9. Cathodic cycling effects in the oxide films formed on zirconium alloys type AB2

    International Nuclear Information System (INIS)

    Zerbino, J.O; Visintin, A; Triaca, W

    2003-01-01

    The passive behavior of ZrNi alloys near the rest potential is studied through in situ voltammetry, ellipsometry, and microscopic observation.A significant oxide layer growth is observed in aqueous 1 M KOH during the application of different potential programs currently used in the activation processes of the alloy.The understanding of both the alloy activation process and the hydrogen absorption process is important in the strategies employed for the design of electrodes for nickel metal hydride batteries.The kinetics of the oxide layer formation, under potential cycling in the cathodic region related to the rest potential, plays a significant role in the activation process of metal alloy.Cathodic potential cycling increases the thickness and decreases the compactness of the passive oxide layer.The protonation of the oxide decreases the barrier effect and makes the anodic polarization more effective.Potential cycling gives rise to increasing surface oxidation, hydrogen absorption and hydride formation, and produces the consequent fragmentation of the material mainly through grain limits (J.Solid State Eletrochem. in press)

  10. Effect of Refiner Addition Level on Zirconium-Containing Aluminium Alloys

    International Nuclear Information System (INIS)

    Jaradeh, M M R; Carlberg, T

    2012-01-01

    It is well known that in aluminium alloys containing Zr, grain refiner additions do not function as desired, producing an effect often referred to as nuclei poisoning. This paper investigates the structure of direct chill-cast ingots of commercial AA3003 aluminium alloys, with and without Zr, at various addition levels of Al5Ti1B master alloy. In Bridgman experiments simulating ingot solidification, Zr-containing alloys were studied after the addition of various amounts of Ti. It could be demonstrated, in both ingot casting and simulation experiments, that Zr poisoning can be compensated for by adding more Ti and/or Al5Ti1B. The results confirm better refinement behaviour with the addition of Ti + B than of only Ti. The various combinations of Zr and Ti also influenced the formation of AlFeMn phases, and the precipitation of large Al 6 (Mn,Fe) particles was revealed. AlZrTiSi intermetallic compounds were also detected.

  11. Effect of Refiner Addition Level on Zirconium-Containing Aluminium Alloys

    Science.gov (United States)

    Jaradeh, M. M. R.; Carlberg, T.

    2012-01-01

    It is well known that in aluminium alloys containing Zr, grain refiner additions do not function as desired, producing an effect often referred to as nuclei poisoning. This paper investigates the structure of direct chill-cast ingots of commercial AA3003 aluminium alloys, with and without Zr, at various addition levels of Al5Ti1B master alloy. In Bridgman experiments simulating ingot solidification, Zr-containing alloys were studied after the addition of various amounts of Ti. It could be demonstrated, in both ingot casting and simulation experiments, that Zr poisoning can be compensated for by adding more Ti and/or Al5Ti1B. The results confirm better refinement behaviour with the addition of Ti + B than of only Ti. The various combinations of Zr and Ti also influenced the formation of AlFeMn phases, and the precipitation of large Al6(Mn,Fe) particles was revealed. AlZrTiSi intermetallic compounds were also detected.

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

    Science.gov (United States)

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

    2000-10-01

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

  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. Fast fracture of a zirconium alloy pressure tube: cause and implications

    International Nuclear Information System (INIS)

    Price, E.G.; Cheadle, B.A.

    1985-12-01

    The cause of the unstable fracture of a Zircaloy-2 pressure tube in the core of a CANDU reactor is reviewed. Failure was associated with the presence of brittle zones of zirconium hydride which developed as a result of thermal gradient induced hydrogen diffusion. Unstable fracture occurred when the partial thickness crack reached an unstable length and the crack ran 2 meters along the tube and terminated by circumferential tearing. The partial thickness defect initiated and propagated to an unstable length by delayed hydride cracking is high compared to fatigue progression and increases exponentially with temperature. Delayed hydride cracking can be prevented by reducing residual stresses to a minimum and by high standards of non-destructive testing that ensures freedom from unacceptable defects. Future prevention of fast fracture is based upon the inspection of a limited number of fuel channels for the presence of defects and for conditions which can cause hydride build-up together with the periodic removal of Zr-2.5wt% Nb tubes to monitor their condition

  15. Criteria for fracture initiation at hydrides in zirconium alloys. Pt. 1

    International Nuclear Information System (INIS)

    Shi, S.Q.; Puls, M.P.

    1994-01-01

    A theoretical framework for the initiation of delayed hydride cracking (DHC) in zirconium is proposed for two different types of initiating sites, i.e., a sharp crack tip (considered in this part) and a shallow notch (considered in part II). In the present part I, an expression for K IH is derived which shows that K IH depends on the size and shape of the hydride precipitated at the crack tip, the yield stress and elastic moduli of the material and the fracture stress of the hydride. If the hydride at the crack tip extends in length at constant thickness, then K IH increases as the square root of the hydride thickness. Thus a microstructure favouring the formation of thicker hydrides at the crack tip would result in an increased K IH . K IH increases slightly with temperature up to a temperature at which there is a more rapid increase. The temperature at which there is a more rapid increase in K IH will increase as the yield stress increases. The model also predicts that an increase in yield stress due to irradiation will cause an overall slight decrease in K IH compared to unirradiated material. There is good agreement between the overall predictions of the theory and experimental results. It is suggested that more careful evaluations of some key parameters are required to improve on the theoretical estimates. (orig.)

  16. Experimental study and modeling of high-temperature oxidation and phase transformation of cladding-tubes made in zirconium alloy

    International Nuclear Information System (INIS)

    Mazeres, Benoit

    2013-01-01

    One of the hypothetical accident studied in the field of the safety studies of Pressurized light Water Reactor (PWR) is the Loss-Of-Coolant-Accident (LOCA). In this scenario, zirconium alloy fuel claddings could undergo an important oxidation at high temperature (T≅ 1200 C) in a steam environment. Cladding tubes constitute the first confinement barrier of radioelements and then it is essential that they keep a certain level of ductility after quenching to ensure their integrity. These properties are directly related to the growth kinetics of both the oxide and the αZr(O) phase and also to the oxygen diffusion profile in the cladding tube after the transient. In this context, this work was dedicated to the understanding and the modeling of the both oxidation phenomenon and oxygen diffusion in zirconium based alloys at high temperature. The numerical tool (EKINOX-Zr) used in this thesis is based on a numerical resolution of a diffusion/reaction problem with equilibrium-conditions on three moving boundaries: gas/oxide, oxide/αZr(O), αZr(O)/βZr. EKINOX-Zr kinetics model is coupled with ThermoCalc software and the Zircobase database to take into account the influence of the alloying elements (Sn, Fe, Cr, Nb) but also the influence of hydrogen on the solubility of oxygen. This study focused on two parts of the LOCA scenario: the influence of a pre-oxide layer (formed in-service) and the effects of hydrogen. Thanks to the link between EKINOX-Zr and the thermodynamic database Zircobase, the hydrogen effects on oxygen solubility limit could be considered in the numerical simulations. Thus, simulations could reproduce the oxygen diffusion profiles measured in pre-hydrided samples. The existence of a thick pre-oxide layer on cladding tubes can induce a reduction of this pre-oxide layer before the growth of a high-temperature one during the high temperature dwell under steam. The first simulations performed using the numerical tool EKINOX-Zr showed that this particular

  17. Novel experiments to characterise creep-fatigue degradation in VHTR alloys

    International Nuclear Information System (INIS)

    Simpson, J.A.; Wright, J.K.; Wright, R.N.

    2015-01-01

    It is well known in energy systems that the creep lifetime of high temperature alloys is significantly degraded when a cyclic load is superimposed on components operating in the creep regime. A test method has been developed in an attempt to characterise creep-fatigue behaviour of alloys at high temperature. The test imposes a hold time during the tensile phase of a fully reversed strain-controlled low cycle fatigue test. Stress relaxation occurs during the strain-controlled hold period. This type of fatigue stress relaxation test tends to emphasise the fatigue portion of the total damage and does not necessarily represent the behaviour of a component in-service well. Several different approaches to laboratory testing of creep-fatigue at 950 deg. C have been investigated for Alloy 617, the primary candidate for application in VHTR heat exchangers. The potential for mode switching in a cyclic test from strain control to load control, to allow specimen extension by creep, has been investigated to further emphasise the creep damage. In addition, tests with a lower strain rate during loading have been conducted to examine the influence of creep damage occurring during loading. Very short constant strain hold time tests have also been conducted to examine the influence of the rapid stress relaxation that occurs at the beginning of strain holds. (authors)

  18. Wettability of zirconium diboride ceramics by Ag, Cu and their alloys with Zr

    International Nuclear Information System (INIS)

    Muolo, M.L.; Ferrera, E.; Novakovic, R.; Passerone, A.

    2003-01-01

    Sintered ZrB 2 ceramics, pure and with 4 wt.% Ni as sintering aid, have been tested in contact with liquid Ag, Cu, Ag-Cu and Ag-Cu-Zr. ''Pure'' ZrB 2 ceramics are wetted by Ag-Zr alloys, and ZrB 2 /Ni ceramics also by pure Cu. The wetting behaviour is briefly discussed in terms of existing wetting theories

  19. 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.

  20. Uranium-zirconium based alloys part I: reference points for thermophysical properties

    International Nuclear Information System (INIS)

    Dias, Marcio Soares; Mattos, Joao Roberto L. de

    2015-01-01

    An integrated modelling process named Relative Variational Model (RVM) is in development by the fuel designers of the CDTN. The lack of measurements in the thermal and physical properties for new fuels, as well as the high dispersion of the existing measurements are challenges in the development of nuclear fuel concepts since that higher uncertainties of the material properties have as result the detrimental reduction on the safety margins . Based on the RVM, the integrated process has been applied to the derivation of reference points for the U-Zr based alloy. (author)

  1. 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

  2. Characterisation of AGR fuel cladding alloy using secondary ion mass spectrometry

    International Nuclear Information System (INIS)

    Allen, G.C.; Sparry, R.P.; Wild, R.K.

    1987-08-01

    Uranium dioxide fuel used in the Advanced Gas Cooled Reactor (AGR) is contained in a ribbed can of 20wt%Cr/25wt%Ni/Nb stabilised steel. Laboratory circumstances, spall during thermal cycling. To date it has been difficult to identify active material originating from the oxidation product of the cladding alloy in the cooling circuit. In an attempt to solve this problem we have set out to characterise fully a sample of oxide from this source and work is in progress to obtain suitable oxide samples from the surface of a 20%Cr/25%Ni/Nb stainless steel. In view of its high sensitivity and the ability to obtain chemical information from relatively small areas we have sought to use Secondary Ion Mass Spectroscopy (SIMS). (author)

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

    Science.gov (United States)

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

    2017-06-01

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

  4. Heterogeneities in metallic glasses. Atomistic computer simulations on the structure and mechanical properties of copper-zirconium alloys and composites

    International Nuclear Information System (INIS)

    Brink, Tobias

    2017-01-01

    The present thesis deals with molecular dynamics computer simulations of heterogeneities in copper-zirconium metallic glasses, ranging from intrinsic structural fluctuations to crystalline secondary phases. These heterogeneities define, on a microscopic scale, the properties of the glass, and an understanding of their nature and behaviour is required for deriving the proper structure-property relations. In terms of composite systems, we start with the amorphisation of copper nanolayers embedded in a metallic glass matrix. While copper is an fcc metal with a high propensity for crystallisation, amorphisation can in fact occur in such systems for thermodynamic reasons. This is due to interface effects, which are also known from heterogeneous interfaces in crystals or from grain boundary complexions, although in absence of lattice mismatch. In single-phase glasses, intrinsic heterogeneities are often discussed in terms of soft spots or geometrically unfavourable motifs (GUMs), which can be considered to be mechanically weaker, defective regions of the glass. We investigate the relation between these motifs and the boson peak, an anomaly in the vibrational spectrum of all glasses. We demonstrate a relation between the boson peak and soft spots by analysing various amorphous and partially amorphous samples as well as highentropy alloys. Finally, we treat the plastic deformation of glasses, with and without crystalline secondary phases. We propose an explanation for the experimentally observed variations of propagation direction, composition, and density along a shear band. These variations of propagation direction are small in the case of single-phase glasses. A considerably greater influence on shear band propagation can be exerted by precipitates. We systematically investigate composites ranging from low crystalline volume fraction up to systems which resemble a nanocrystalline metal. In this context, we derive a mechanism map for composite systems and observe the

  5. Capsule development and utilization for material irradiation tests; study on the in-pile creep measuring method of zirconium alloys

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Yong; Lee, Byung Kee; Lee, Jong Jea; Kim, Chang Sik; Kim, B. Hun; Cho, I. Sik [Sunmoon University, Asan (Korea)

    2002-02-01

    The final objective of this project is to obtain a design and fabrication technology of an in-pile creep test machine of zirconium alloys. First, design concepts of the in-pile creep test machines of various foreign countries were reviewed and a preliminary design of the equipment was carried. Second, the mock-up of the in-pile creep test machine was fabricated based on the preliminary design. The mock-up consisted of upper and lower grips, a yoke, a pressure chamber including a bellows, a push rod and LVDT. Each part was made of 304 L stainless steel. The average surface roughness of the parts was 1.0-14.7 {mu}m. The mock-up precisely determined an extension of a specimen by gas pressure. Finally, in-pile creep capsule was designed, fabricated and modified. High pure aluminum blocks were put in the capsule. Considering heat transfer coefficients of helium and nitrogen gases, the cooling efficiency is about 4 .deg. C at the condition of 300 .deg. C creep test. Yield strength, ultimate tensile strength and elongation at 300 .deg. C were 335 MPa, 591 MPa, 19.8%, respectively. which were lower than the values at room temperature, 353 MPa, 740 MPa, 12.5%. This study gave an important technology related to design, fabrication and performance tests of the in-pile creep test machine, which is applied to the fabrication of a special capsule and also used for the fundamental data for the fabrication of various in-pile creep capsules. 6 refs., 45 figs., 5 tabs. (Author)

  6. Heterogeneities in metallic glasses. Atomistic computer simulations on the structure and mechanical properties of copper-zirconium alloys and composites

    Energy Technology Data Exchange (ETDEWEB)

    Brink, Tobias

    2017-07-01

    The present thesis deals with molecular dynamics computer simulations of heterogeneities in copper-zirconium metallic glasses, ranging from intrinsic structural fluctuations to crystalline secondary phases. These heterogeneities define, on a microscopic scale, the properties of the glass, and an understanding of their nature and behaviour is required for deriving the proper structure-property relations. In terms of composite systems, we start with the amorphisation of copper nanolayers embedded in a metallic glass matrix. While copper is an fcc metal with a high propensity for crystallisation, amorphisation can in fact occur in such systems for thermodynamic reasons. This is due to interface effects, which are also known from heterogeneous interfaces in crystals or from grain boundary complexions, although in absence of lattice mismatch. In single-phase glasses, intrinsic heterogeneities are often discussed in terms of soft spots or geometrically unfavourable motifs (GUMs), which can be considered to be mechanically weaker, defective regions of the glass. We investigate the relation between these motifs and the boson peak, an anomaly in the vibrational spectrum of all glasses. We demonstrate a relation between the boson peak and soft spots by analysing various amorphous and partially amorphous samples as well as highentropy alloys. Finally, we treat the plastic deformation of glasses, with and without crystalline secondary phases. We propose an explanation for the experimentally observed variations of propagation direction, composition, and density along a shear band. These variations of propagation direction are small in the case of single-phase glasses. A considerably greater influence on shear band propagation can be exerted by precipitates. We systematically investigate composites ranging from low crystalline volume fraction up to systems which resemble a nanocrystalline metal. In this context, we derive a mechanism map for composite systems and observe the

  7. 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

  8. Simultaneous multielement analysis of zirconium alloys by chlorination separation of matrix/ICP-AES

    International Nuclear Information System (INIS)

    Kato, Kaneharu

    1990-01-01

    An analytical method combined chlorination separation of matrix with ICP-AES has been developed for reactor grade Zr alloys (Zircaloy-2). A sample (1 g) is taken into a Pt boat and chlorinated with HCl gas of 100 ml/min in a glass reaction tube at ca. 330degC. Matrix Zr of the sample is volatilized and separated as ZrCl 4 . The analytic elements remaining quantitatively as chlorination residue are dissolved in a mixture of mineral acids (6 M HCl 3 ml+conc. HNO 3 0.5 ml+conc. H 2 SO 4 0.2 ml) and diluted to 20 ml with distilled water after filtration. ICP-AES was used for simultaneous multielement determination using a calibration curve method. The present method has the following advantages: simple sample preparation procedure; applicability to any form of samples to determine multielements; simple ICP-AES calibration procedure. This method was successfully applied to the determination of Fe, Ni, Cu, Co, Mn and Pb in the Zr alloys of JAERI CRM's and NBS SRM's. (author)

  9. Hydride-induced degradation of hoop ductility in textured zirconium-alloy tubes: A theoretical analysis

    International Nuclear Information System (INIS)

    Qin, W.; Szpunar, J.A.; Kozinski, J.

    2012-01-01

    Hydride-induced degradation of hoop ductility in Zr-alloy tubular components has been studied for many years because of its importance in the nuclear industry. In this paper the role of intergranular and intragranular δ-hydrides in the degradation of ductility of the textured Zr-alloy tubes is investigated. The correlation among hydride distribution, orientation and morphology in the tubes is formulated based on thermodynamic modeling, and then analyzed. The results show that the applied stress, the crystallographic texture of α-Zr matrix, the grain-boundary structure, and the morphology and size of Zr grains simultaneously govern the site preference and the orientation of hydrides. A criterion is proposed to determine the threshold stress of hydride reorientation. The hoop ductility of the hydrided Zr tubes is discussed using the concept of macroscopic fracture strain. It is shown that the intergranular hydrides may be more deleterious to ductility than the intragranular ones. This work defines a general framework for understanding the relation of the microstructure of hydride-forming materials to embrittlement.

  10. Zirconium and cast zirconium

    Energy Technology Data Exchange (ETDEWEB)

    Krone, K

    1977-04-01

    A survey is given on the occurence of zirconium, production of Zr sponge and semi-finished products, on physical and mechanical properties, production of Zr cast, composition of the commercial grades and reactor grades qualities, metal cutting, welding, corrosion behavior and use.

  11. 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)

  12. Indoor atmospheric corrosion of historical ferrous alloys. System characterisation, mechanisms and modelling discussion

    International Nuclear Information System (INIS)

    Monnier, J.

    2008-12-01

    Understanding the mechanisms of indoor atmospheric corrosion in iron alloys is of primary importance in several fields, including for the conservation of Middle Ages monuments or the long term storage of nuclear waste. In this research, a double approach was developed, combining fine characterisation of corrosion systems and design of experiments to answers specific questions related to mechanisms understanding. Iron indoor atmospheric corrosion was investigated on samples coming from the reinforcing chain of the Amiens cathedral (15. century). In the first stage, the corrosion system has been extensively characterised from the macroscopic to the nano-metric scale. In particular, structural micro-analysis (μ-Raman, μ-XRD, μ-XAS) has been used to locate, identify and quantify the oxidised phases. Rust layers are composed of a matrix of nano-metric goethite, with low quantities of lepidocrocite and akaganeite mostly located in the extern part of the corrosion system. In addition, clear marblings are dispersed in the matrix, which are sometimes connected with the metal core. Although these may contain maghemite, these marblings are generally made of ferri-hydrite/feroxyhite phases. In the second stage, specific experiments have been carried out in an unsaturated marked medium to locate oxygen reduction sites in the rust layers. Several cases were evidenced, depending on the rust layer morphology. In addition, reduction processes of model phases have been studied in situ, using an electrochemical cell coupled with structural characterisation techniques. This combination highlighted the influence of reduction mode and pH on the type of reduced phase formed. From the obtained results, several mechanisms are proposed to explain the long term indoor atmospheric corrosion of iron, including rust layers morphology and phases properties. The different hypotheses have been integrated in a proposed method to diagnosis ancient ferrous systems stability. These hypotheses also

  13. Evolution of interphase and intergranular strain in zirconium-niobium alloys during deformation at room temperature

    Science.gov (United States)

    Cai, Song

    Zr-2.5Nb is currently used for pressure tubes in the CANDU (CANada Deuterium Uranium) reactor. A complete understanding of the deformation mechanism of Zr-2.5Nb is important if we are to accurately predict the in-reactor performance of pressure tubes and guarantee normal operation of the reactors. This thesis is a first step in gaining such an understanding; the deformation mechanism of ZrNb alloys at room temperature has been evaluated through studying the effect of texture and microstructure on deformation. In-situ neutron diffraction was used to monitor the evolution of the lattice strain of individual grain families along both the loading and Poisson's directions and to track the development of interphase and intergranular strains during deformation. The following experiments were carried out with data interpreted using elasto-plastic modeling techniques: (1) Compression tests of a 100%betaZr material at room temperature. (2) Tension and compression tests of hot rolled Zr-2.5Nb plate material. (3) Compression of annealed Zr-2.5Nb. (4) Cyclic loading of the hot rolled Zr-2.5Nb. (5) Compression tests of ZrNb alloys with different Nb and oxygen contents. The experimental results were interpreted using a combination of finite element (FE) and elasto-plastic self-consistent (EPSC) models. The phase properties and phase interactions well represented by the FE model, the EPSC model successfully captured the evolution of intergranular constraint during deformation and provided reasonable estimates of the critical resolved shear stress and hardening parameters of different slip systems under different conditions. The consistency of the material parameters obtained by the EPSC model allows the deformation mechanism at room temperature and the effect of textures and microstructures of ZrNb alloys to be understood. This work provides useful information towards manufacturing of Zr-2.5Nb components and helps in producing ideal microstructures and material properties for

  14. Morphology and hot deformation of lamellar microstructures in zirconium and titanium alloys

    International Nuclear Information System (INIS)

    Vanderesse, N.

    2008-06-01

    This study aims at providing a precise description of the lamellar microstructures of two alloys, Zircaloy-4 and TA6V, and at characterizing their deformation at high temperature. New experimental techniques have been developed for these materials: instrumented Jominy end quench test, channel-die with mobile walls, X-ray microtomography. The main results underline the role of the alpha-GB phase formed at the prior beta grain boundaries on the variant selection in Zircaloy-4 and TA6-V. The three dimensional organization of the colonies in TA6V is also revealed for the first time and discussed in relationship with the formation of the microstructure. In hot compressed Zircaloy-4 several mechanisms of strain localization are observed. Twinning activity at 750 C, in particular, is clearly put into evidence. A classification of these heterogeneities is proposed and their influence on the recrystallization is discussed. (author)

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

    Science.gov (United States)

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

    2016-07-01

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

  16. 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

  17. Results from the characterisation of the Futurix-FTA metal alloy transmutation fuels

    International Nuclear Information System (INIS)

    Rory Kennedy, J.; O'Holleran, Th.; Keiser, D.

    2007-01-01

    Full text of publication follows. Idaho National Laboratory has been developing and irradiation testing a number of fuels and fuel types for actinide transmutation as part of the Advanced Fuel Cycle Initiative (AFCI). Fuel types under consideration include both fertile (fast reactor systems) and fertile-free (accelerator-driven systems) metallic alloys. Most recently, fuel fabrication was completed and the fuel pins shipped to the fast flux Phenix reactor in Marcoule, France for irradiation testing as part of the FUTURIX-FTA experiment: an international experiment involving the USA, France, the European Commission and Japan. The metal alloy fuels for this experiment are the low-fertile U-29Pu-4Am-2Np-30Zr and the non-fertile Pu-12Am-40Zr. The fresh fuels have been fully characterised for chemical composition, phase, microstructure, thermal behaviour and fuel-cladding-chemical-interaction (FCCI). Preliminary FCCI results raised some safety concerns with respect to the formation of low melting phases and cladding degradation, which could preclude a fuel from consideration. Results from diffusion couple experiments between the non-fertile fuel Pu-12Am-40Zr and the ferritic HT9 and 422 stainless steels (SS) used in the AFC experiments in the ATR reactor (USA) compared to the austenitic AIM1 SS used in the FUTURIX-FTA experiments in the Phenix reactor (France) indicate significant inter-diffusion with the AIM1 SS. Up to about a 30-fold increase in the diffusion of iron (and accompanying Ni and Cr) into the fuel at 650 C was observed compared to the 422 SS studies. Comparable studies between the low-fertile U-29Pu-4Am-2Np-30Zr fuel alloy and the AIM1 SS show virtually no inter-diffusion. The Fe (along with small amounts of Ni and Cr) appears as small precipitates in the fuel alloy with only minor concentrations identified in the fuel alloy matrix. These results will be discussed in terms of mechanisms of the inter-diffusion and the difference in behaviour between the

  18. Evolution of microstructure in zirconium alloy core components of nuclear reactors during service

    International Nuclear Information System (INIS)

    Griffiths, M.; Coleman, C.E.; Holt, R.A.; Sagat, S.; Urbanic, V.F.; Chow, C.K.

    1993-03-01

    X-ray diffraction and analytical electron microscopy have been used to characterise microstructural and microchemical changes produced by neutron irradiation of Zr-2.5Nb, Zircaloy-2 and Zircaloy-4 nuclear reactor core components. In many cases there is a clear relationship between the radiation damage microstructure and the physical properties of in-service core components. For example, the difference in delayed hydride cracking velocity between the inlet and outlet ends of Zr-2.5Nb pressure tubes in pressurised heavy water reactors can be directly correlated with variations in a-dislocation density and β-Zr phase decomposition. For the same tubes, the variation of fracture toughness has the same fluence dependence as dislocation loop density and improvements in corrosion behaviour can be linked with decreases in the Nb concentration in the α-Zr matrix due to Nb precipitation during irradiation. For pressurised water reactors and boiling water reactors the onset of 'breakaway' growth in Zircaloy-4 guide tubes can be directly correlated with the appearance of basal plane dislocation loops in the microstructure. (author). 37 refs., 28 figs., 4 tabs

  19. Evolution of microstructure in zirconium alloy core components of nuclear reactors during service

    Energy Technology Data Exchange (ETDEWEB)

    Griffiths, M; Coleman, C E; Holt, R A; Sagat, S; Urbanic, V F [Atomic Energy of Canada Ltd., Chalk River, ON (Canada). Chalk River Nuclear Labs.; Chow, C K [Atomic Energy of Canada Ltd., Pinawa, MB (Canada). Whiteshell Nuclear Research Establishment

    1993-03-01

    X-ray diffraction and analytical electron microscopy have been used to characterise microstructural and microchemical changes produced by neutron irradiation of Zr-2.5Nb, Zircaloy-2 and Zircaloy-4 nuclear reactor core components. In many cases there is a clear relationship between the radiation damage microstructure and the physical properties of in-service core components. For example, the difference in delayed hydride cracking velocity between the inlet and outlet ends of Zr-2.5Nb pressure tubes in pressurised heavy water reactors can be directly correlated with variations in a-dislocation density and {beta}-Zr phase decomposition. For the same tubes, the variation of fracture toughness has the same fluence dependence as dislocation loop density and improvements in corrosion behaviour can be linked with decreases in the Nb concentration in the {alpha}-Zr matrix due to Nb precipitation during irradiation. For pressurised water reactors and boiling water reactors the onset of `breakaway` growth in Zircaloy-4 guide tubes can be directly correlated with the appearance of basal plane dislocation loops in the microstructure. (author). 37 refs., 28 figs., 4 tabs.

  20. Radiation damage of austenitic stainless steels and zirconium alloys; Pregled radijacionog ostecenja austenitnih nerdjajucih celika i legura cirkonijuma

    Energy Technology Data Exchange (ETDEWEB)

    Stefanovic, V [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Yugoslavia)

    1965-11-15

    This review contains analyses of available data concerning texture deformations and radiation damage of zirconium and zircaloy-2; radiation damage, influence of neutron radiation on the mechanical properties of austenitic, ferritic and other types of stainless steels.

  1. 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)

  2. 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

  3. Mechanical behaviour and failure of fuel cladding zirconium alloys in nuclear power plants under accidental RIA-type situation

    International Nuclear Information System (INIS)

    Doan, D.T.

    2009-01-01

    In French Nuclear Pressurized Water Reactors (PWRs), most of structural parts of the fuel assembly consist of zirconium alloy tubes and plates. Optimizing the management of fuel in nuclear power plants led to the increase in the duration of fuel cycles and power. The use of high fuel burnups requires drastic changes in the rules for reactor design in the nuclear safety. The evaluation of nuclear reactors in accident situations is based on reference accident scenarios. One of these hypothetical accidents, examined in this study, is the 'Reactivity Initiated Accident'. In order to assess the structural integrity of these parts it is necessary to characterize both the plastic flow and fracture behaviour of the materials at various stages of the life cycle, (i.e. at increasing levels of hydriding, irradiation, oxidation or thermal mechanical loading). The purpose of this work is to provide experimental data and to develop a model of the thermo-mechanical behaviour and to propose a design analysis method in the case of non-irradiated clads, in RIA-type situations. Mechanical tests were conducted on Cold-Worked-Stress-Relieved and on Recrystallized Zircaloy-4 sheets using various kinds of samples including smooth and notched tensile specimens and small punch tests. Temperature was set to 25, 250 and 600 C with hydrogen contents between 0 and 1000 ppm. The model is based on a simplified description of a Zircaloy polycrystal in which scalar isotropic ductile damage including void nucleation and growth is added. The model is also physically based to easily transfer parameters determined for one material state to another (e.g. transfer between sheet and tube or between different levels of irradiation). The model was implemented in the Finite Element software Zebulon using either an explicit or an implicit time integration scheme. Uniaxial tension tests were used to tune the model parameters for both materials, considering various values of temperature and hydrogen levels

  4. 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)

  5. 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

  6. Characterisation of lead-calcium alloys ageing in anisothermal conditions by calorimetric, resistance and hardness in-situ measurements

    Energy Technology Data Exchange (ETDEWEB)

    Rossi, F.; Lambertin, M. [Arts et Metiers ParisTech, LaBoMaP, Cluny (France); Delfaut-Durut, L. [CEA, Centre de Valduc, (SEMP, LECM), Is-sur-Tille (France); Maitre, A. [SPCTS, UFR Sciences et techniques, Limoges (France); Vilasi, M. [LCSM, Univ. Nancy I, Vandoeuvre les Nancy (France)

    2010-02-15

    Transformations undergone by lead-calcium alloys are numerous and have different kinetics from a few minutes to a few years. Anisothermal calorimetric measurements were performed to be able to identify these transformations quickly. It was then possible to identify five transformations. Complementary measurements have enabled us to define transformations with an in-situ electrical resistance measurement to follow the evolution of the calcium in solid solution and with an in-situ hardness measurement to characterise the mechanical properties. The aim of these results is to simulate the ageing and overageing of the alloy in isothermal conditions. (orig.)

  7. The surface oxidation kinetics of zirconium-niobium alloys and aα-Fe with prevailing cubical texture

    International Nuclear Information System (INIS)

    Mukhambetov, D.G.; Kargin, D.B.; Chalaya, O. V.; Berber, N.N.

    2002-01-01

    It is known, that the kinetics of oxidation of zirconium at formed heating is characterized by two consecutive stages. At the initial stage the thin protecting film will be derived. The relation of its depth from time h (t) is described predominantly by parabolic law. Some time later there can be a transition to the linear law of oxidation. The time moment divided these areas on the kinetic relation is called as a point of break. The film is formed at the second stage, has a developed grid of pores or cracks, can be flake away and be crumbled by losing its protective properties. At the oxidation of the surface shells of the heat generating elements and the technological channels of atomic boilers both stages are proceeded simultaneously. This phenomenon is called modular corrosion. Its consequences can be dangerous for the equipment. Its mechanism is not clear till now. Similar dependencies h(t), with the break point, beginning from which the thin film is transformed into the thick one were found by us at the oxidation α-Fe with prevailing cubical texture. The task of the work was to study the oxide film growth laws in order to clarify the mechanisms of transition of the thin film into the oxide layer on the α-Fe surface and Zr-Nb alloy modular corrosion emergence. Low-carbonate steel with contents 99.43 % of α-Fe was used as a model object of our research. In the texture of the steel surface planar direction [100] was prevalent. Its part accounted for about 40 %. The isothermal air oxidation was carried out in the interval of 450-500 deg. C . Phase composition of the film was determined with X-ray diffraction. The mathematical treatment of the dependencies h(t) obtained by experiment showed that the kinetics of the film growth can be conditionally divided into 4-stages. The initial stage is described by function logarithmic function, the other stages - by the power mode h n =A n ·t, namely, the second stage - is described by function close to cubical (n≅3

  8. TEM characterisation of stress corrosion cracks in nickel based alloys: effect of chromium content and chemistry of environment

    International Nuclear Information System (INIS)

    Delabrouille, F.

    2004-11-01

    Stress corrosion cracking (SCC) is a damaging mode of alloys used in pressurized water reactors, particularly of nickel based alloys constituting the vapour generator tubes. Cracks appear on both primary and secondary sides of the tubes, and more frequently in locations where the environment is not well defined. SCC sensitivity of nickel based alloys depends of their chromium content, which lead to the replacement of alloy 600 (15 % Cr) by alloy 690 (30 % Cr) but this phenomenon is not yet very well understood. The goal of this thesis is two fold: i) observe the effect of chromium content on corrosion and ii) characterize the effect of environment on the damaging process of GV tubes. For this purpose, one industrial tube and several synthetic alloys - with controlled chromium content - have been studied. Various characterisation techniques were used to study the corrosion products on the surface and within the SCC cracks: SIMS; TEM - FEG: thin foil preparation, HAADF, EELS, EDX. The effect of chromium content and surface preparation on the generalised corrosion was evidenced for synthetic alloys. Moreover, we observed the penetration of oxygen along triple junctions of grain boundaries few micrometers under the free surface. SCC tests show the positive effect of chromium for contents varying from 5 to 30 % wt. Plastic deformation induces a modification of the structure, and thus of the protective character, of the internal chromium rich oxide layer. SCC cracks which developed in different chemical environments were characterised by TEM. The oxides which are formed within the cracks are different from what is observed on the free surface, which reveals a modification of medium and electrochemical conditions in the crack. Finally we were able to evidence some structural characteristics of the corrosion products (in the cracks and on the surface) which turn to be a signature of the chemical environment. (author)

  9. 9-12% Cr heat resistant steels. Alloy design, TEM characterisation of microstructure evolution and creep response at 650 C

    Energy Technology Data Exchange (ETDEWEB)

    Rojas Jara, David

    2011-03-21

    This work was carried out aiming to design and characterise 9-12% Cr steels with tailormade microstructures for applications in fossil fuel fired power plants. The investigations concentrated in the design and characterisation of heat resistant steels for applications in high oxidising atmospheres (12% Cr) and 9% Cr alloys for components such as rotors (P91). ThermoCalc calculations showed to be a reliable tool for alloy development. The modeling also provided valuable information for the adjustment of the processing parameters (austenisation and tempering temperatures). Two 12% Cr heat resistant steels with a fine dispersion of nano precipitates were designed and produced supported by thermodynamic modeling (ThermoCalc). A detailed characterisation of the microstructure evolution at different creep times (100 MPa / 650 C / 8000 h) was carried out by scanning transmission electron microscopy (STEM). The results of the microstructure analysis were correlated with the mechanical properties in order to investigate the influence of different precipitates (especially M{sub 23}C{sub 6} carbides) on the creep strength of the alloys. Precipitation of Laves phase and Z-phase was observed after several hundred hours creep time. Very few Z-phase of the type Cr(V,Ta)N nucleating from existing (V,Ta)(C,N) was observed. Both alloys show growth and coarsening of Laves phase, meanwhile the MX carbonitrides present a very slow growth and coarsening rate. Alloys containing Laves phase, MX and M{sub 23}C{sub 6} precipitates show best creep properties. The influence of hot-deformation and tempering temperature on the microstructure evolution on one of the designed 12% Cr alloys was studied during short-term creep at 80-250 MPa and 650 C. Quantitative determination of dislocation density and sub-grain size in the initial microstructure and after creep was investigated by STEM combined with the high-angle annular dark-field detector (HAADF). A correlation between microstructure

  10. 9-12% Cr heat resistant steels. Alloy design, TEM characterisation of microstructure evolution and creep response at 650 C

    International Nuclear Information System (INIS)

    Rojas Jara, David

    2011-01-01

    This work was carried out aiming to design and characterise 9-12% Cr steels with tailormade microstructures for applications in fossil fuel fired power plants. The investigations concentrated in the design and characterisation of heat resistant steels for applications in high oxidising atmospheres (12% Cr) and 9% Cr alloys for components such as rotors (P91). ThermoCalc calculations showed to be a reliable tool for alloy development. The modeling also provided valuable information for the adjustment of the processing parameters (austenisation and tempering temperatures). Two 12% Cr heat resistant steels with a fine dispersion of nano precipitates were designed and produced supported by thermodynamic modeling (ThermoCalc). A detailed characterisation of the microstructure evolution at different creep times (100 MPa / 650 C / 8000 h) was carried out by scanning transmission electron microscopy (STEM). The results of the microstructure analysis were correlated with the mechanical properties in order to investigate the influence of different precipitates (especially M 23 C 6 carbides) on the creep strength of the alloys. Precipitation of Laves phase and Z-phase was observed after several hundred hours creep time. Very few Z-phase of the type Cr(V,Ta)N nucleating from existing (V,Ta)(C,N) was observed. Both alloys show growth and coarsening of Laves phase, meanwhile the MX carbonitrides present a very slow growth and coarsening rate. Alloys containing Laves phase, MX and M 23 C 6 precipitates show best creep properties. The influence of hot-deformation and tempering temperature on the microstructure evolution on one of the designed 12% Cr alloys was studied during short-term creep at 80-250 MPa and 650 C. Quantitative determination of dislocation density and sub-grain size in the initial microstructure and after creep was investigated by STEM combined with the high-angle annular dark-field detector (HAADF). A correlation between microstructure evolution and creep

  11. Elastoplastic phase-field modeling of ζ-hydride precipitation in zirconium alloy: dynamics evolution in inhomogeneous elasticity

    International Nuclear Information System (INIS)

    Oum, G.; Thuinet, L.; Legris, A.

    2015-07-01

    A phase-field (PF) model was developed within the framework of homogeneous and heterogeneous elasticity theory to study the precipitation of ζ-hydride in zirconium. By coupling crystal plasticity to PF we show that plastic strain participates in lowering the transformation stresses, and therefore induces changes in nucleation, growth and morphology evolution of the precipitates. (authors)

  12. Delayed Hydride Cracking Mechanism in Zirconium Alloys and Technical Requirements for In-Service Evaluation of Zr-2.5Nb Tubes with Flaws

    International Nuclear Information System (INIS)

    Kim, Young Suk

    2007-01-01

    In association with periodic inspection of CANDU nuclear power plant components, Canadian Standards Association issued CSA N285.8 in 2005 as technical requirements for in-service evaluation of zirconium alloy pressure tubes in CANDU reactors. This first version, CSA N285.8 involves procedures for, firstly, the evaluation of pressure tube flaws, secondly, the evaluation of pressure tube to calandria tube contact and, thirdly, the assessment of a reactor core, and material properties and derived quantities. The evaluation of pressure tube flaws includes delayed hydride cracking evaluation the procedures of which are stipulated based on the existing delayed hydride cracking models. For example, the evaluation of flaw-tip hydride precipitation during reactor cooldown involves a procedure to calculate the equilibrium hydrogen equivalent concentration in solution at the flaw tip, Htipas follows: Htip=Hfexp[- (VH delta no.)/RT], where Hf is the total bulk hydrogen equivalent concentration, VH partial molar volume of hydrogen in zirconium, δ a difference in hydrostatic stress between the bulk and the crack tip. When Htip ≥TSSP at temperature, then flaw-tip hydride is predicted to precipitate. Eq. (1) suggests that hydrogen concentration at the crack tip would increase due to an work energy given by the difference in the hydrostatic stress

  13. 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

  14. 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

  15. 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.

  16. 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

  17. 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

  18. Properties of zirconium silicate and zirconium-silicon oxynitride high-k dielectric alloys for advanced microelectronic applications: Chemical and electrical characterizations

    Science.gov (United States)

    Ju, Byongsun

    2005-11-01

    As the microelectronic devices are aggressively scaled down to the 1999 International Technology Roadmap, the advanced complementary metal oxide semiconductor (CMOS) is required to increase packing density of ultra-large scale integrated circuits (ULSI). High-k alternative dielectrics can provide the required levels of EOT for device scaling at larger physical thickness, thereby providing a materials pathway for reducing the tunneling current. Zr silicates and its end members (SiO2 and ZrO2) and Zr-Si oxynitride films, (ZrO2)x(Si3N 4)y(SiO2)z, have been deposited using a remote plasma-enhanced chemical vapor deposition (RPECVD) system. After deposition of Zr silicate, the films were exposed to He/N2 plasma to incorporate nitrogen atoms into the surface of films. The amount of incorporated nitrogen atoms was measured by on-line Auger electron spectrometry (AES) as a function of silicate composition and showed its local minimum around the 30% silicate. The effect of nitrogen atoms on capacitance-voltage (C-V) and leakage-voltage (J-V) were also investigated by fabricating metal-oxide-semiconductor (MOS) capacitors. Results suggested that incorporating nitrogen into silicate decreased the leakage current in SiO2-rich silicate, whereas the leakage increased in the middle range of silicate. Zr-Si oxynitride was a pseudo-ternary alloy and no phase separation was detected by x-ray photoelectron spectroscopy (XPS) analysis up to 1100°C annealing. The leakage current of Zr-Si oxynitride films showed two different temperature dependent activation energies, 0.02 eV for low temperature and 0.3 eV for high temperature. Poole-Frenkel emission was the dominant leakage mechanism. Zr silicate alloys with no Si3N4 phase were chemically separated into the SiO2 and ZrO2 phase as annealed above 900°C. While chemical phase separation in Zr silicate films with Si 3N4 phase (Zr-Si oxynitride) were suppressed as increasing the amount of Si3N4 phase due to the narrow bonding network m Si3

  19. 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; Quantification de la repartition de l'hydrogene a la microsonde nucleaire du Laboratoire Pierre Sue dans l'epaisseur de tubes de gainage du combustible des REP en alliage de zirconium

    Energy Technology Data Exchange (ETDEWEB)

    Raepsaet, C. [CEA Saclay, Dept. de Recherche sur l' Etat Condense, les Atomes et les Molecules (DSM/DRECAM/LPS-CNRS) UMR9956, 91 - Gif sur Yvette (France); Bossis, Ph. [CEA Saclay, Dept. des Materiaux pour le Nucleaire (DEN/DANS/DMN/SEMULM2E), 91 - Gif-sur-Yvette (France); Hamon, D.; Bechade, J.L.; Brachet, J.C. [CEA Saclay, Dept. des Materiaux pour le Nucleaire (DEN/DANS/DMN/SRMALA2M), 91 - Gif-sur-Yvette (France)

    2007-07-01

    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.)

  20. Synthesis and Properties of Metallic Technetium and Technetium-Zirconium Alloys as Transmutation Target and Radioactive waste storage form in the UREX+1 Process

    Energy Technology Data Exchange (ETDEWEB)

    Hartmann, Thomas [Idaho State University/Idaho National Laboratory, 1776 Science Center Drive, Idaho Falls, ID 83402 (United States)]|[Harry Reid Center, University Nevada - Las Vegas, 4505 Maryland Parkway, Las Vegas, NV (United States); Poineau, Frederic; Czerwinski, Kenneth R. [Harry Reid Center, University Nevada - Las Vegas, 4505 Maryland Parkway, Las Vegas, NV (United States)

    2008-07-01

    In the application of UREX+1 process, technetium will be separated together with uranium and iodine within the first process step. After the separation of uranium, technetium and iodine must be immobilized by their incorporation in a suitable waste storage-form. Based on recent activities within the AFCI community, a potential candidate as waste storage form to immobilize technetium is to alloy the metal with excess zirconium. Alloys in the binary Tc-Zr system may act as potential transmutation targets in order to transmute Tc-99 into Ru-100. We are presenting first results in the synthesis of metallic technetium, and the synthesis of equilibrium phases in the binary Tc-Zr system at 1400 deg. C after arc-melting and isothermal annealing under inert conditions. Samples were analyzed using X-ray powder diffraction, Rietveld analysis, scanning electron microscopy, and electron probe micro-analysis, which allows us to construct the binary Tc-Zr phase diagram for the isothermal section at 1400 deg. C. (authors)

  1. Elaboration and characterisation of Pd-Cr alloys for PEM fuel cells

    Energy Technology Data Exchange (ETDEWEB)

    Souleymane, B.; Fouda-Onana, F.; Savadogo, O. [Ecole Polytechnique de Montreal, Montreal, PQ (Canada). Laboratoire de nouveaux materiaux pour l' energie et l' electrochimie

    2008-07-01

    Palladium (Pd) alloys have been considered as alternative catalyst cathodes for the oxygen reduction reaction (ORR) in proton exchange membrane (PEM) fuel cells, particularly in liquid fuel cells. The purpose of this study was to investigate the ORR on various Pd-Cr alloys. Pd-Cr alloys were deposited on glassy carbon support and the electrocatalytic parameters for the ORR were determined in acid medium. The effect of the Pd-Cr alloy deposition parameters on its composition and electrocatalytic behaviour were determined. The study showed that there is a relationship between the composition of the alloy and the power of the Pd and Cr cathode. The parameters of the ORR were correlated to the alloy chemical and physical properties. EDS and XPS analysis revealed a segregation of Cr in the alloy.The variation of the work function (W) of the alloy with the alloy composition has shown a minimum value of W of 0.287 for a composition of the alloy of 70 per cent of Pd and 30 per cent of Cr. The electrochemically active surface area and the exchange current density of the ORR indicated that the mechanism of the ORR on Pd-Cr is similar to that on platinum. 9 refs., 2 figs.

  2. 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

  3. Characterization of the active deformation mechanisms in Zirconium alpha alloys, and use of micro-macro transfer models

    International Nuclear Information System (INIS)

    Francillette, H.; Bacroix, B.; Gasperini, M.; Lebensohn, R.A.

    1996-01-01

    The aim of this study is to model the evolution of the crystallographic textures of rolled zirconium sheet metals, based on the active deformation mechanisms. Plane compression tests have been carried out on Zr 702 polycrystalline samples, at ambient temperature. Active mechanisms were identified and characterized by the means of local orientation measurements (EBSD: electron BackScattering Diffraction), completed with global texture measurements. Measured orientations are then introduced in Taylor, Sachs and self-coherent type micro-macro models in order to validate these models with respect to mechanism activation and texture evolution. (A.B.)

  4. 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, ...

  5. Modelling, characterisation and uncertainties of stabilised pseudoelastic shape memory alloy helical springs

    DEFF Research Database (Denmark)

    Enemark, Søren; Santos, Ilmar; Savi, M. A.

    2016-01-01

    The thermo-mechanical behaviour of pseudoelastic shape memory alloy helical springs is of concern discussing stabilised and cyclic responses. Constitutive description of the shape memory alloy is based on the framework developed by Lagoudas and co-workers incorporating two modifications related t...

  6. 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

  7. Microstructural, mechanical characterisation and fractography of As-cast Ti-Al alloy

    International Nuclear Information System (INIS)

    Hamzah, E.; Ong, W.R.; Tamin, M.N.

    2007-01-01

    The effect of alloying element, namely chromium (Cr) on the microstructures, mechanical characterization and fracture surface of gamma titanium aluminide (Ti Al) has been studied. Micro-hardness and fatigue crack growth tests were performed on as-cast samples with composition of Ti-48at%Al and Ti-48%Al-2at%Cr. Prior to the micro-hardness tests; samples were metallurgically prepared for microstructural and structural analysis using optical microscope and scanning electron microscope. Field emission scanning electron microscope (FESEM) technique was employed to investigate the fracture surface of sample after fatigue crack growth test. Micro-hardness tests results showed increasing hardness value of Ti-48Al alloys when chromium is added. Both titanium aluminide alloys exhibited a nearly lamellae microstructure. However, finer laths of plates in lamellar structure have been observed in Ti-48at%Al-2at%Cr. FESEM micrograph of surface fracture indicates a mixed mode of failure for both alloys. (author)

  8. Morphological characterisation and spectroscopic studies of the corrosion behaviour of tungsten heavy alloys

    International Nuclear Information System (INIS)

    Ogundipe, A.; Greenberg, B.; Braida, W.; Christodoulatos, C.; Dermatas, D.

    2006-01-01

    Tungsten-based alloys have been used in a wide variety of industrial and military applications. These alloys are composed mainly of tungsten (88-95%) with various combinations of nickel, cobalt, iron and copper usually making up the remaining fraction. The corrosion behaviours of five munitions grade tungsten alloys of interest have been examined using immersion tests and wet-dry cycle tests to determine the mechanisms involved in the release of the metallic components. Analyses carried out using SEM, EDS and grazing incidence XRD techniques, show the release of tungsten as well as alloying elements due to galvanic corrosion resulting from the difference in electrode potential between the tungsten phase and the binder phase in all cases studied. The extent of corrosion was directly related with the dissolution of tungsten in the binder phase during the sintering stage of manufacture. In W-Ni-Co-Fe alloys binder phase corrosion was observed while the relatively noble tungsten phase was less affected. The reverse was observed for a W-Cu alloy

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

    Energy Technology Data Exchange (ETDEWEB)

    Filippov, V. P., E-mail: vpfilippov@mephi.ru; Bateev, A. B.; Lauer, Yu. A. [National Research Nuclear University “MEPhI” (Moscow Engineering Physics Institute) (Russian Federation)

    2016-12-15

    Mössbauer investigations were used to compare iron atom states in oxide films of binary Zr-Fe, ternary Zr-Fe-Cu and quaternary Zr-Fe-Cr-Sn alloys. Oxide films are received in an autoclave at a temperature of 350–360 °C and at pressure of 16.8 MPa. The corrosion process decomposes the intermetallic precipitates in alloys and forms metallic iron with inclusions of chromium atoms α–Fe(Cr), α–Fe(Cu), α–Fe {sub 2}O{sub 3} and Fe {sub 3}O{sub 4} compounds. Some iron ions are formed in divalent and in trivalent paramagnetic states. The additional doping influences on corrosion kinetics and concentration of iron compounds and phases formed in oxide films. It was shown the correlation between concentration of iron in different chemical states and corrosion resistance of alloys.

  10. 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.

  11. Characterization of adiabatic shear bands in the zirconium alloy impacted by split Hopkinson pressure bar at a strain rate of 6000 s−1

    International Nuclear Information System (INIS)

    Zou, D.L.; Luan, B.F.; Liu, Q.; Chai, L.J.; Chen, J.W.

    2012-01-01

    The adiabatic shear bands formed in the zirconium alloy impacted by split Hopkinson pressure bar at a strain rate of about 6000 s −1 were characterized systemically by means of a high resolution field emission scanning electron microscope equipped with electron backscatter diffraction probe. The results show that the transformed bands were distinguished on the cross-section view of the impacted specimens, and the ultrafine and equiaxed grains formed in the transformed bands were confirmed. The gradient variation of the grains across the transformed bands from the boundary to the center of the bands was observed, and the grains at the center of the transformed bands were finer than other zones. Based on the characterization of the deformed microstructure adjacent to the transformed bands, the formation mechanism of the ultrafine and equiaxed grains in the transformed bands was revealed, and the rotational dynamic recrystallization mechanism should be responsible for the formation of the ultrafine and equiaxed grains in the transformed bands. According to the collection of the cumulative misorientation at different strain levels, the formation and evolution process of the ultrafine and equiaxed grains in the transformed bands were speculated. The microhardness measurements show that high microhardness value in the transformed bands was obtained because of the grain refining, and the large standard deviation of the microhardness at the center of the transformed bands was confirmed due to the gradient microstructural distribution in the bands.

  12. A self-consistent anisotropic approach for the simulation of plastic deformation and texture development of polycrystals: Application to zirconium alloys

    International Nuclear Information System (INIS)

    Lebensohn, R.A.; Tome, C.N.

    1993-01-01

    The authors present in this work a visco-plastic self-consistent (VPSC) anisotropic approach for modeling the plastic deformation of polycrystals, together with a thorough discussion of the assumptions involved and the range of application of such approach. They use the VPSC model for predicting texture development during rolling and axisymmetric deformation of zirconium alloys, and to calculate the yield locus and the Lankford coefficient of rolled Zircaloy sheet. They compare the results with experimental data and find that they are in good agreement with the available experimental evidence. They also compare the VPSC prediction with the ones of a Full Constraints approach and observe that they differ both quantitatively and qualitatively: according with the predictions of the VPSC scheme, deformation is accommodated mostly by the soft systems, the twinning activity is much lower, and fewer systems are active, in average, per grain. These results are a consequence of having accounted for the grain interaction with its surroundings, which is a crucial aspect when modeling plastically anisotropic materials

  13. 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.)

  14. Microstructure and Mechanical Properties of Nano-Size Zirconium Carbide Dispersion Strengthened Tungsten Alloys Fabricated by Spark Plasma Sintering Method

    International Nuclear Information System (INIS)

    Xie Zhuoming; Liu Rui; Fang Qianfeng; Zhang Tao; Jiang Yan; Wang Xianping; Liu Changsong

    2015-01-01

    W-(0.2, 0.5, 1.0)wt% ZrC alloys with a relative density above 97.5% were fabricated through the spark plasma sintering (SPS) method. The grain size of W-1.0wt% ZrC is about 2.7 μm, smaller than that of pure W and W-(0.2, 0.5)wt% ZrC. The results indicated that the W-ZrC alloys exhibit higher hardness at room temperature, higher tensile strength at high temperature, and a lower ductile to brittle transition temperature (DBTT) than pure W. The tensile strength and total elongation of W-0.5wt% ZrC alloy at 700 °C is 535 MPa and 24.8%, which are respectively 59% and 114% higher than those of pure W (337 MPa, 11.6%). The DBTT of W-(0.2, 0.5, 1.0)wt% ZrC materials is in the range of 500°C–600°C, which is about 100 °C lower than that of pure W. Based on microstructure analysis, the improved mechanical properties of the W-ZrC alloys were suggested to originate from the enhanced grain boundary cohesion by ZrC capturing the impurity oxygen in tungsten and nano-size ZrC dispersion strengthening. (paper)

  15. 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)

  16. Effect of zirconium on grain growth and mechanical properties of a ball-milled nanocrystalline FeNi alloy

    Energy Technology Data Exchange (ETDEWEB)

    Kotan, Hasan, E-mail: hkotan@ncsu.edu [Department of Materials Science and Engineering, NC State University, 911 Partners Way, Room 3078, Raleigh, NC 27606-7907 (United States); Darling, Kris A. [U.S. Army Research Laboratory, Weapons and Materials Research Directorate, RDRL-WMM-F, Aberdeen Proving Ground, MD 21005-5069 (United States); Saber, Mostafa; Koch, Carl C.; Scattergood, Ronald O. [Department of Materials Science and Engineering, NC State University, 911 Partners Way, Room 3078, Raleigh, NC 27606-7907 (United States)

    2013-02-25

    Highlights: Black-Right-Pointing-Pointer Pure Fe, Fe{sub 92}Ni{sub 8}, and Fe{sub 91}Ni{sub 8}Zr{sub 1} powders were hardened up to 10 GPa by ball milling. Black-Right-Pointing-Pointer Annealing of Fe and Fe{sub 92}Ni{sub 8} leads to reduced hardness and extensive grain growth. Black-Right-Pointing-Pointer The addition of Zr to Fe{sub 92}Ni{sub 8} increases its stability and strength by second phases. Black-Right-Pointing-Pointer The second phases are found to promote the stability of Fe{sub 91}Ni{sub 8}Zr{sub 1} by Zener pinning. Black-Right-Pointing-Pointer The Zr-containing precipitates contribute to the overall strength of the material. - Abstract: Grain growth of ball-milled pure Fe, Fe{sub 92}Ni{sub 8}, and Fe{sub 91}Ni{sub 8}Zr{sub 1} alloys has been studied using X-ray diffractometry (XRD), focused ion beam (FIB) microscopy and transmission electron microscopy (TEM). Mechanical properties with respect to compositional changes and annealing temperatures have been investigated using microhardness and shear punch tests. We found the rate of grain growth of the Fe{sub 91}Ni{sub 8}Zr{sub 1} alloy to be much less than that of pure Fe and the Fe{sub 92}Ni{sub 8} alloy at elevated temperatures. The microstructure of the ternary Fe{sub 91}Ni{sub 8}Zr{sub 1} alloy remains nanoscale up to 700 Degree-Sign C where only a few grains grow abnormally whereas annealing of pure iron and the Fe{sub 92}Ni{sub 8} alloy leads to extensive grain growth. The grain growth of the ternary alloy at high annealing temperatures is coupled with precipitation of Fe{sub 2}Zr. A fine dispersion of precipitated second phase is found to promote the microstructural stability at high annealing temperatures and to increase the hardness and ultimate shear strength of ternary Fe{sub 91}Ni{sub 8}Zr{sub 1} alloy drastically when the grain size is above nanoscale.

  17. 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

  18. 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

  19. Characterisation and modelling of vacancy dynamics in Ni–Mn–Ga ferromagnetic shape memory alloys

    Energy Technology Data Exchange (ETDEWEB)

    Merida, D., E-mail: david.merida@ehu.es [Fisika Aplikatua II Saila, Euskal Herriko Unibertsitatea UPV/EHU, p.k. 644, 48080 Bilbao (Spain); Elektrizitate eta Elektronika Saila, Euskal Herriko Unibertsitatea UPV/EHU, p.k. 644, 48080 Bilbao (Spain); García, J.A. [Fisika Aplikatua II Saila, Euskal Herriko Unibertsitatea UPV/EHU, p.k. 644, 48080 Bilbao (Spain); BC Materials (Basque Centre for Materials, Application and Nanostructures), 48040 Leioa (Spain); Sánchez-Alarcos, V. [Departamento de Física, Universidad Pública de Navarra, Campus de Arrosadia, 31006 Pamplona (Spain); Pérez-Landazábal, J.I.; Recarte, V. [Departamento de Física, Universidad Pública de Navarra, Campus de Arrosadia, 31006 Pamplona (Spain); Institute for Advanced Materials (INAMAT), Universidad Pública de Navarra, Campus de Arrosadía, 31006 Pamplona (Spain); Plazaola, F. [Elektrizitate eta Elektronika Saila, Euskal Herriko Unibertsitatea UPV/EHU, p.k. 644, 48080 Bilbao (Spain)

    2015-08-05

    Highlights: • We study the dynamics of vacancies for three different Ni–Mn–Ga alloy samples. • The formation and migration energies have been obtained experimentally. • The entropic factor and the distance a vacancy has to reach a sink are measured. • We present a theoretical model to explain the dynamics of vacancies. • Results are applicable for any thermal treatment and extensible to other alloys. - Abstract: The dynamics of vacancies in Ni–Mn–Ga shape memory alloys has been studied by positron annihilation lifetime spectroscopy. The temperature evolution of the vacancy concentration for three different Ni–Mn–Ga samples, two polycrystalline and one monocrystalline, have been determined. The formation and migration energies and the entropic factors are quite similar in all cases, but vary slightly according to composition. However, the number of jumps a vacancy has to overtake to reach a sink is five times higher in the single crystal. This is an expected result, due to the role that surfaces and grain boundaries should play in balancing the vacancy concentration. In all cases, the initial vacancy concentration for the samples quenched from 1173 K lies between 1000 ppm and 2000 ppm. A phenomenological model able to explain the dynamics of vacancies has been developed in terms of the previous parameters. The model can reproduce the vacancy dynamics for any different kind of thermal history and can be easily extended to other alloys.

  20. Long-term durability test of acid recovery evaporators made of Ti-5% Ta alloy and zirconium

    International Nuclear Information System (INIS)

    Takeuchi, Masayuki; Koizumi, Tsutomu; Koyama, Tomozo

    2001-05-01

    Mock-ups of acid recovery evaporators which are made of Ti-5% Ta alloy and Zr were tested under inactive condition for forty thousands hours to improve a corrosion resistance of acid recovery evaporator in Tokai reprocessing plant (TRP). The mock-up unit was designed and produced referring to the specification of acid recovery evaporator in TRP and the evaporation performance of the mock-up was 1/27 of TRP. A long-term durability of both evaporators was demonstrated by results of operation data, evaporation performance and corrosion resistance. The mock-up unit did not suffer from any trouble during the running test and the operation data such as temperature, flow, concentrations of nitric acid and metal ions were fairly stable within standard condition. As for the corrosion resistance, cracks and local corrosion such as intergranular attack were not observed on both evaporators after the running test, and a corrosion of weld was not selective. The average corrosion rates at measuring points were less than 0.1 mm/yr, respectively, however, thickness of the Ti-5% Ta alloy evaporator was slightly reduced at all points of vapor phase region. In addition, from the result by test coupon, it is found that both materials have low susceptibility to stress corrosion cracking in this environment. The destructive inspection showed that the mechanical properties of both materials were not degraded during the running test. Finally, the total running time of the mock-up unit is much more than a maximum running time of acid recovery evaporator made of stainless steel in TRP (nearly 15,000 hours). On the basis of the test results, an excellent durability of Ti-5% Ta alloy and Zr evaporators under was successfully demonstrated throughout the mock-up test from an engineering perspective. (author)

  1. Application of cast nickel alloys for parts of electronics characterised by special magnetic properties

    Directory of Open Access Journals (Sweden)

    W. UhI

    2008-03-01

    Full Text Available Thc thcorctical part of the study highlights thc origin of thc idca 10 start investigations on alloys of high ~nngnctic pcrmcability.manufactured mainly by cornpanics in ~ h Uc S A and Japan.'Phc said materials arc applicd for various pans of ctcctronics uscd by thc military industry. c.g. sntctlitc antcnnas Tor globalcommunication with suhmarincs. and for rcscarch instmrncnts, c,g. fcrromagnctic corcs. Thcy arc chnr:~clcriscd by vcry high lnnpncticpcrrncability. resistivity and corrosion rcsistancc which makc thcm suitablc for opcrat ion undcr cxtrn-~ryingc onditions.Nickel alloys of high magnctic propcrtics arc usuall y manufactured as roZlcd products. The amhition of t hc authors or this srlldy is Inmanufacture !hem as cast prnducts.Thc pmgram of rcscarch incIudcd characteristic of nickcl alloys wirh ddi t i ons of molybdcnum slid iron sn~isryingt hc ahnvc mc~iito ncdrcquircmcnu. with attcn~ionf ocusscd on thcir application for magnctic parts of satcllitc antcnnns and fcrromngnctic corcs.Moulding and casting tcchnologics wcrc proposcd to bcst suit ~ h pcr occss OF maaufacturc of r hcsc clcmcnrs.Thc rangc of chcmicaI cornpostion was sclcctcd 20 cnsurc thc rcquircd magnctic. mcchnnicnl and anti-corrosive pmpcrtics.A scrics of melts was prcparcd and castings of thc abovc mcnlioncd clclncn1s wcrc mndc. Thc chclnicnl composi~ioii of IEIC alloys wasanalyscd along with thc stnlcturc cxarninations nnd quality asscssmcnt rnadc by ~ h cno n-dcsrructi vc rncthods, Casrings wcrc sitbjcctcd tothc finishing trcatmcnt, followed by tests and cxamina~ionsto cnablc thcir practical application.

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

    International Nuclear Information System (INIS)

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

    1983-01-01

    Superconducting upper critical fields H/sub c/2(T), transition temperatures T/sub c/, and normal-state electrical resistivities rho/sub n/ have been measured in the amorphous transition-metal alloy series Zr/sub 1-z/Co/sub x/, Zr/sub 1-x/Ni/sub x/, (Zr/sub 1-x/Ti/sub x/)/sub 0.78/Ni/sub 0.22/, and (Zr/sub 1-x/Nb/sub x/)/sub 0.78/Ni/sub 0.22/. Structural integrity of these melt-spun alloys is indicated by x-ray, density, bend-ductility, normal-state electrical resistivity, superconducting transition width, and mixed-state flux-pinning measurements. The specimens display T/sub c/ = 2.1--3.8 K, rho/sub n/ = 159--190 μΩ cm, and Vertical Bar(dH/sub c/2/dT)cVertical Bar = 28--36 kG/K. These imply electron mean free paths lroughly-equal2--6 A, zero-temperature Ginzburg-Landau coherence distances xi/sub G/0roughly-equal50--70 A, penetration depths lambda/sub G/0roughly-equal(7--10) x 10 3 A, and extremely high dirtiness parameters xi 0 /lroughly-equal300--1300. All alloys display H/sub c/2(T) curves with negative curvature and (with two exceptions) fair agreement with the standard dirty-limit theory of Werthamer, Helfand, Hohenberg, and Maki (WHHM) for physically reasonable values of spin-orbit-coupling induced, electron-spin-flip scattering time tau/sub so/. This is in contrast to the anomalously elevated H/sub c/2(T) behavior which is nearly linear in T that is observed by some, and the unphysically low-tau/sub so/ fits to WHHM theory obtained by others, for various amorphous alloys

  3. Impurity composition effect on work function in cylindrical specimens of niobium and low zirconium niobium base alloys

    International Nuclear Information System (INIS)

    Kobyakov, V.P.

    2000-01-01

    A study is made into poly- and single crystal cylindrical niobium specimens, prepared by various methods as well as into polycrystalline specimens of niobium base alloys doped with 1.2 and 1.6 % Zr. Thermionic work function is measured using a full current method. Several techniques are applied to determine the content of substitutional and interstitial impurities in specimens. The phase composition of polished section surface is also investigated. A work function increase is observed when a considerable amount of carbide phases occurs at the surface. This increase is comparable with the effect of going from a polycrystalline niobium specimen to a single crystal with (110) surface orientation [ru

  4. Corrosion mechanisms of zirconium alloys - study of the initial oxidation kinetics and of the mechanical behaviour of the metal/oxide system

    International Nuclear Information System (INIS)

    Parise, M.

    1996-12-01

    Nuclear fuel claddings are made of zirconium alloys. The conditions of use lead the cladding oxidize outside. The so-formed layers behaves like a thermal barrier and prevents from using oxidized claddings with an oxide thickness larger than 100 μm. The oxidation kinetic is approximately cubic for oxide thicknesses smaller than about 2μm, linear beyond. A kinetic model has been proposed which estimates the post-transition growth rate from the kinetic parameters of the pre-transition state and morphological features of post-transition layers. This work aims at providing the necessary elements to validate this model and studying the layers around the kinetic transition, in order to determine whether the oxidation mechanisms before and after the transition are similar. Thicknesses of the 50 - 500 nm range of the oxide layers are measured by an optical method; pre-transition kinetics are thus precisely determined. The effect of the composition, the thermal treatment and the presence of oxygen in solid solution is studied. The morphological and crystallographic study of the layers show that they exhibit a lot of similarities before and after the kinetic transition. The results concerning the kinetic aspects and the morphology of the post-transition layers point out that the proposed model leads to realistic post-transition growth rates. Furthermore, the kinetic transition corresponds to the appearance of cracks in the oxide layer. The mechanical behaviour of the metal/oxide system has been modelled at different scales. When the specific behaviours of the metal and the oxide are taken into account together with the interface geometry, radial stresses appear, which are high enough to locally open cracks. The appearance and localization of cracks depend on both the interface geometry and the stress distribution in the metal/oxide system. (author)

  5. Evaluation of Conditions for Hydrogen Induced Degradation of Zirconium Alloys during Fuel Operation and Storage. Final Report of a Coordinated Research Project 2011-2015

    International Nuclear Information System (INIS)

    2015-12-01

    This publication reports on the work carried out in 2011–2015 in the coordinated research project (CRP) on the evaluation of conditions for hydrogen induced degradation of zirconium alloys during fuel operation and storage. The CRP was carried out to evaluate the threshold condition for delayed hydride cracking (KIH) in pressurized water reactors and zircaloy-4 and E635M fuel claddings, with application to in-pile operation and spent fuel storage. The project consisted of adding hydrogen to samples of cladding and measuring K IH by one of four methods. The CRP was the third in the series, of which the results of the first two were published in IAEA-TECDOC-1410 and IAEA-TECDOC-1649, in 2004 and 2010, respectively. This publication includes all of the research work performed in the framework of the CRP, including details of the experimental procedures that led to a set of data for tested materials. The research was conducted by representatives from 13 laboratories from all over the world. In addition to the basic goal to transfer the technology of the testing techniques from experienced laboratories to those unfamiliar with the methods, the CRP was set up to develop experimental procedures to produce consistent sets of data, both within a single laboratory and among different laboratories. The material condition and temperature history were prescribed, and laboratories chose one or two of four methods of loading that were recommended in an attempt to develop standard sets of experimental protocols so that consistent results could be obtained. Experimental discrepancies were minimized through careful attention to details of microstructure, temperature history and stress state in the samples, with the main variation being the mode of loading

  6. Silver- and Zirconium-added ternary and quaternary TiAu based high temperature shape memory alloys

    Energy Technology Data Exchange (ETDEWEB)

    Wadood, A., E-mail: abdul.wadood@ist.edu.pk [High Temperature Materials Unit, National Institute for Materials Science, 1-2-1 Sengen, Tsukuba, Ibaraki 305-0047 (Japan); Department of Materials Science and Engineering, Institute of Space Technology (IST), Near Rawat Toll Plaza, Islamabad (Pakistan); Yamabe-Mitarai, Y. [High Temperature Materials Unit, National Institute for Materials Science, 1-2-1 Sengen, Tsukuba, Ibaraki 305-0047 (Japan)

    2015-10-15

    Low strength in B2 phase, incomplete shape memory effect and high cost of Au are obstacles for the use of Ti–50Au as a high temperature shape memory alloy. We investigated the effects of partial substitution of Ti with Zr and Au with Ag in Ti–Au on phase constitution, phase transformation, and high temperature thermo-mechanical and shape memory properties. Partial substitution of Ti with Zr in Ti–50Au and Ti–40Au–10Ag was found to improve the thermo-mechanical and shape memory effect. However, partial substitution of Au with Ag in Ti–50Au and Ti–50Au–10Zr was found to have negligible effects. Reasons for such different behavior of Zr- and Ag-added Ti–Au alloys are considered. - Highlights: • Au, Ag and Ti, Zr belong to same group. Effects of partial substitution of Au with Ag and Ti with Zr in Ti–Au are investigated. • Zr was found more effective than Ag in improving shape memory and mechanical properties. • Same atomic size of Au and Ag and large size misfit b/w Ti and Zr atoms. • Ag resulted large amount of precipitation in Ti–Au.

  7. Performance evaluation and characterisation of EIGA produced titanium alloy powder for additive manufacturing processes

    CSIR Research Space (South Africa)

    Arthur, Nana KK

    2017-11-01

    Full Text Available affect powder quality, and hinder processing. In an investigation by Goso and Kale [3], Ti-6Al-4V alloy powder was produced by the hydride-dehydride (HDH) process in order to make titanium components by blended elemental approach. Chemical analysis.... 2016. Additive manufacturing of metals, Acta Materialia, 117, pp 371-392. 3 [3] Goso, X. and Kale, A. 2010. Production of titanium metal powder by the HDH process, (Paper presented at the South African Institute of Mining and Metallurgy Light...

  8. Some microstructural characterisations in a friction stir welded oxide dispersion strengthened ferritic steel alloy

    International Nuclear Information System (INIS)

    Legendre, F.; Poissonnet, S.; Bonnaillie, P.; Boulanger, L.; Forest, L.

    2009-01-01

    The goal of this study is to characterize microstructure of a friction stir welded oxide dispersion strengthened alloy. The welded material is constituted by two sheets of an yttria-dispersion-strengthened PM 2000 ferritic steel. Different areas of the friction stir welded product were analyzed using field emission gun secondary electron microscopy (FEG-SEM) and electron microprobe whereas nanoindentation was used to evaluate mechanical properties. The observed microstructural evolution, including distribution of the yttria dispersoids, after friction stir welding process is discussed and a correlation between the microstructure and the results of nanoindentation tests is established.

  9. 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

  10. Microstructural and micromechanical characterisation of TiAl alloys using atomic force microscopy and nanoindentation

    International Nuclear Information System (INIS)

    Gebhard, S.; Pyczak, F.; Goeken, M.

    2009-01-01

    Different microstructures were generated in the Ti-45Al-4.6Nb-0.2B-0.2C and Ti-45Al-1Cr alloys (at.%) by heat treatment. The microstructures were investigated using nanoindentation and atomic force microscopy which was compared with transmission electron microscopy. Topographic contrast is usually used for phase identification in the atomic force microscope. However, it was found that the topographic order of the phases changes with different microstructures and specimen preparations. Nanoindentation measurements provided local hardness values not obtainable by other methods and enabled clear distinction of the phases. The hardness values can give information on surrounding microstructure and solid solution hardening. The mean lamellar spacing of the colonies was measured using both atomic force microscopy and transmission electron microscopy. Atomic force microscopy was found to be suitable to determine the spacing between α 2 /γ-interfaces offering the advantages of easier sample preparation and fewer specimens compared to evaluation by TEM analysis.

  11. Experimental characterisation and modelling of deformation- induced microstructure in an A6061 aluminium alloy

    International Nuclear Information System (INIS)

    Kreyca, J F; Falahati, A; Kozeschnik, E

    2016-01-01

    For industry, the mechanical properties of a material in form of flow curves are essential input data for finite element simulations. Current practice is to obtain flow curves experimentally and to apply fitting procedures to obtain constitutive equations that describe the material response to external loading as a function of temperature and strain rate. Unfortunately, the experimental procedure for characterizing flow curves is complex and expensive, which is why the prediction of flow-curves by computer modelling becomes increasingly important. In the present work, we introduce a state parameter based model that is capable of predicting the flow curves of an A6061 aluminium alloy in different heat-treatment conditions. The model is implemented in the thermo-kinetic software package MatCalc and takes into account precipitation kinetics, subgrain formation, dynamic recovery by spontaneous annihilation and dislocation climb. To validate the simulation results, a series of compression tests is performed on the thermo-mechanical simulator Gleeble 1500. (paper)

  12. Experimental characterisation and modelling of deformation- induced microstructure in an A6061 aluminium alloy

    Science.gov (United States)

    Kreyca, J. F.; Falahati, A.; Kozeschnik, E.

    2016-03-01

    For industry, the mechanical properties of a material in form of flow curves are essential input data for finite element simulations. Current practice is to obtain flow curves experimentally and to apply fitting procedures to obtain constitutive equations that describe the material response to external loading as a function of temperature and strain rate. Unfortunately, the experimental procedure for characterizing flow curves is complex and expensive, which is why the prediction of flow-curves by computer modelling becomes increasingly important. In the present work, we introduce a state parameter based model that is capable of predicting the flow curves of an A6061 aluminium alloy in different heat-treatment conditions. The model is implemented in the thermo-kinetic software package MatCalc and takes into account precipitation kinetics, subgrain formation, dynamic recovery by spontaneous annihilation and dislocation climb. To validate the simulation results, a series of compression tests is performed on the thermo-mechanical simulator Gleeble 1500.

  13. Characterisation of phases in nanostructured, multilayered titanium alloys by analytical and high-resolution electron microscopy.

    Science.gov (United States)

    Czyrska-Filemonowicz, A; Buffat, P A

    2009-01-01

    Surface processing of a Ti-6Al-4V alloy led to a complex multilayered microstructure containing several phases of the Ni-Ti-P-Al-O system, which improves the mechanical and tribological surface properties. The microstructure, chemical and phase compositions of the hard layer formed on the surface were investigated by LM, XRD, SEM as well as analytical/high-resolution TEM, STEM, EDS, electron diffraction and FIB. Phase identification based on electron diffraction, HRTEM and EDS microanalysis revealed the presence of several binary and ternary phases in the system Ti-Ni-P, sometimes with partial substitution of Ti by Al. However some phases, mainly nanoparticles, still remain not identified satisfactorily. Electron microscopy techniques used for identification of phases present in surface multilayers and some practical limits to their routine application are reminded here.

  14. Formation of a cerium conversion coating on magnesium alloy using ascorbic acid as additive. Characterisation and anticorrosive properties of the formed films

    OpenAIRE

    A.P. Loperena; I.L. Lehr; S.B. Saidman.

    2016-01-01

    Cerium-based conversion coatings were formed on AZ91D magnesium alloy by immersion of the substrate in solutions containing Ce(NO3)3, H2O2 and ascorbic acid (HAsc). The characterisation of the films was performed by electrochemical and surface analysis techniques such as SEM, EDS, X-ray diffraction and X-ray photoelectron spectroscopy (XPS). The degree of corrosion protection achieved was evaluated in simulated physiological solution by the open circuit potential monitoring, polarisation tech...

  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. Non-destructive characterisation of polymers and Al-alloys by polychromatic cone-beam phase contrast tomography

    International Nuclear Information System (INIS)

    Kastner, Johann; Plank, Bernhard; Requena, Guillermo

    2012-01-01

    X-ray computed tomography (XCT) has become a very important tool for the non-destructive characterisation of materials. Continuous improvements in the quality and performance of X-ray tubes and detectors have led to cone-beam XCT systems that can now achieve spatial resolutions down to 1 μm and even below. Since not only the amplitude but also the phase of an X-ray beam is altered while passing through an object, phase contrast effects can occur even for polychromatic sources when the spatial coherence due to a small focal spot size is high enough. This can lead to significant improvements over conventional attenuation-based X-ray computed tomography. Phase contrast can increase by edge enhancement the visibility of small structures and of features which are only slightly different in attenuation. We report on the possibilities of polychromatic cone-beam phase contrast tomography for non-destructive characterisation of materials. A carbon fibre-reinforced polymer and the Al-alloys AlMg5Si7 and AlSi18 were investigated with high resolution cone-beam X-ray computed tomography with a polychromatic tube source. Under certain conditions strong phase contrast resulting in an upward and downward overshooting of the grey values across edges was observed. The phase effects are much stronger for the polymer than for the Al-alloys. The influence on the phase contrast of the parameters, including source-detector distance, focal spot size and tube acceleration voltage is presented. Maximum phase contrast was observed for a maximum distance between the source and the detector, for a low voltage and a minimum focal spot size at the X-ray source. The detectability of the different phases is improved by the edge enhancement and the resulting improvement of sharpness. Thus, a better segmentation of the carbon fibres in the fibre-reinforced polymer and of the Mg 2 Si-phase in the AlMg5Si7-alloy is achieved. Primary and eutectic Si cannot be detected by attenuation-based X

  17. Texture characterisation of hexagonal metals: Magnesium AZ91 alloy, welded by laser processing

    International Nuclear Information System (INIS)

    Kouadri, A.; Barrallier, L.

    2006-01-01

    Cooled and cast magnesium AZ91 alloy was welded using a CO 2 laser. The changes in the microstructure were analysed by optical and scanning electron microscopy and X-ray diffraction. Modification of the anisotropic properties was evaluated by the characterization of the texture in the base metal, in the core of the welded zone and in the welded zone close to the surface. In the two former zones, we have not observed a texture. Laser welding only leads to a change of the grain size and a disappearance of the eutectic phase. By contrast, in the welded zone close to the surface, the laser process leads both to a finer microstructure, to a loss of the Al-content and to the presence of several texture components. In this zone, our results showed that these textures are on pyramidal {101-bar 1} and prismatic {101-bar 0} planes. Much of the explanation for such texture rests with the fact that during the laser welding, material solidifies in strong non-equilibrium conditions. The kinetics of the nucleation and the growth are partly controlled by the high-rise and high fall of the temperature and the power produced by the laser process. The nature of the texture has been explained by the presence of a columnar to equiaxed transition in the welded zone

  18. Structural and corrosion characterization of hydroxyapatite/zirconium nitride-coated AZ91 magnesium alloy by ion beam sputtering

    Energy Technology Data Exchange (ETDEWEB)

    Kiahosseini, Seyed Rahim, E-mail: rkiahoseyni@yahoo.com [Young Researchers and Elite Club, Damghan Branch, Islamic Azad University, Damghan (Iran, Islamic Republic of); Afshar, Abdollah [Department of Material Science and Engineering, Sharif University of Technology, Tehran (Iran, Islamic Republic of); Mojtahedzadeh Larijani, Majid [Radiation Applications Research School, Nuclear Science and Technology Research Institute, Tehran (Iran, Islamic Republic of); Yousefpour, Mardali [Faculty of Materials and Metallurgical Engineering, Semnan University, Semnan, 35131-19111 (Iran, Islamic Republic of)

    2017-04-15

    Highlights: • The thickness of HA coatings increase by ion beam sputtering time. • The residual strain in HA structure decrease by deposition time increment. • Crystallite size of HA coatings increase by deposition time increment. • The best corrosion resistance occurs at intermediate deposition time. - Abstract: The adhesion of hydroxyapatite (HA) as a coating for the AZ91 magnesium alloy substrate can be improved by using the sputtering method and an intermediate layer, such as ZrN. In this study, HA coatings were applied on ZrN intermediate layers at a temperature of 300 °C for 180, 240, 300, 360, and 420 min by ion beam sputtering. A profilometer device was used to study the HA coating thickness, which changed from 2 μm for the 180-min deposition to 4.7 μm for 420-min deposition. The grazing incidence X-ray diffraction analysis method and the Williamson–Hall analysis were used for structural investigation. As the deposition time increased, the crystalline size increased from 50 nm to 690 nm. However, given sufficient time for stress relief on the coating structure, the lattice strain values were close to zero. Energy-dispersive X-ray spectroscopy results showed that the Ca/P ratio ranged from 1.73 to 1.81. The external indentation method was used to evaluate the coating adhesion to the substrate. The slope of curve for applied force changes versus the radius of cracks in the coating (dP/dr) varied in the range of 0.2–0.07 by the deposition time, indicating that the adhesion increased with the increase in coating thickness. The potentiodynamic polarization technique was used to study the corrosion behavior. With increasing deposition time, the corrosion potential of samples did not show a significant change, and the corrosion potential of all samples (coated and uncoated substrates) was more positive than approximately 55 mV. When the deposition time increased to 360 min, the corrosion current density decreased from 5.5 μA/cm{sup 2} to 0.33

  19. 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)

  20. 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)

  1. 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)

  2. 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

  3. Corrosion and characterisation of dual phase Mg–Li–Ca alloy in Hank’s solution: The influence of microstructural features

    International Nuclear Information System (INIS)

    Zeng, Rong-Chang; Sun, Lu; Zheng, Yu-Feng; Cui, Hong-Zhi; Han, En-Hou

    2014-01-01

    Highlights: •A schematic four-layered structural model of the natural oxide film has been constructed. •A novel concept for the Pilling–Bedworth ratios of chemical compounds is proposed. •Grain refinement in the microstructure leads to a shift from pitting corrosion to overall corrosion. •A method to characterise the corrosion rate of dual phase Mg–Li–Ca alloys is proposed. -- Abstract: The influence of the microstructure and the oxide film of the Mg–9.29Li–0.88Ca alloy on its corrosion behaviour was investigated using SEM, EPMA, XPS and corrosion measurements. The results demonstrated that the fine-grained microstructure improved the mechanical and corrosion resistance of the alloy and shifted pitting corrosion to overall corrosion. The oxide film contained a multi-layered structure, with the outer layer being enriched in lithium-bearing compounds; the interior layer predominantly consisting of oxides, hydroxides and carbonates of lithium and magnesium; and the bottom layer containing oxides. The Pilling–Bedworth ratio for chemical compounds was proposed, and the corrosion rates were characterised

  4. Evaluation of methods for mathematical corrections in the determination of niobium and zirconium contents in U-Nb and U-Zr alloys by X-ray fluorescence analysis

    International Nuclear Information System (INIS)

    Salvador, V.L.R.; Sato, I.M.; Lordello, A.R.

    1985-01-01

    Methods for the determination of niobium and zirconium in U-Nb and U-Zr alloys with the X-ray fluorescence technique are described. The NbK sub(Ab) line, although not under the overlapping effect of the uranium lines as the NbK sub(β) line, although not under the overlapping effect of the uranium lines as the NbK sub(α) is, presents a more intensive absorption effect than this last one; on the other hand the ZrK sub(α) and ZrK sub(β) lines are under the overlapping effect of the uranium spectrum. Such interferences are mathematically corrected by means of relations between the intensities of the lines for the elements and those for the uranium. The technique for the preparation of the samples is the double layer pressed pellet. From the different corrections the best method has showed a precision of 5%. (Author) [pt

  5. Characterisation of intermetallic particles formed during solution treatment of an Al–7Si–0.4Mg–0.12Fe alloy

    International Nuclear Information System (INIS)

    Yao, J.Y.; Taylor, J.A.

    2012-01-01

    Highlights: ► It is concluded in this work that the fine needle-like particles formed during the solution treatment of the 356 type alloy are indeed β phase plates similar to those frequently observed in as-cast aluminium alloys but of much finer sizes, most likely formed by precipitation in the vicinity of the dissolving π phase particles. ► The crystal structure of the needle-like particles, as examined by selected area electron diffraction, can be essentially indexed as either the A-Centred Monoclinic lattice or the Orthorhombic lattice . ► The particles appear to co-exist in two distinct forms: single crystal particles and composite micro-constituent particles. - Abstract: When Fe-containing Al–7Si–xMg alloys are solution-treated, usually as the initial stage of a T6 heat treatment, some of the π-Al 9 Mg 3 FeSi 5 phase present in the as-cast alloy transforms into fine needle-like intermetallic particles (as seen on a polished surface). These precipitated needles, speculated to be β-Al 5 FeSi phase, have not been definitively characterised to date. In this work, electron microscopy characterisation by various techniques was conducted to assess the nature (i.e. the composition, morphology and crystallography) of these particles to verify or otherwise the validity of the above hypothesis. It is found that the particles are indeed β phase platelets, of the same Al 5 FeSi or Al 4.5 FeSi stoichiometry as particles formed during solidification. Close observation of their crystallographic structure suggests fine-scale internal complexities in some of the particles.

  6. 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.

  7. Microstructural characterisation of high-entropy alloy AlCoCrFeNi fabricated by laser engineered net shaping

    Energy Technology Data Exchange (ETDEWEB)

    Kunce, I., E-mail: ikunce@wat.edu.pl [Department of Advanced Materials and Technology, Military University of Technology, 2 Kaliskiego Str., 01-908 Warsaw (Poland); Polanski, M.; Karczewski, K. [Department of Advanced Materials and Technology, Military University of Technology, 2 Kaliskiego Str., 01-908 Warsaw (Poland); Plocinski, T.; Kurzydlowski, K.J. [Faculty of Materials Science and Engineering, Warsaw University of Technology, 141 Wołoska Str., 02-507 Warsaw (Poland)

    2015-11-05

    Laser engineered net shaping (LENS) was used to produce thin-walled samples of the high-entropy alloy AlCoCrFeNi from a prealloyed powder. To determine the effect of the cooling rate during solidification on the microstructure of the alloy, different laser scanning rates were used. A microstructural study of the surfaces of the sample walls was performed using X-ray diffraction analysis and optical and scanning/transmission electron microscopy. The crystal structure of the alloy was determined to be a body-centred cubic (bcc)-derivative B2-ordered type. The microstructure of the alloy produced by LENS was dendritic. Further, it was found that with an increase in the laser scanning rate from 2.5 to 40 mm s{sup −1}, the average grain size decreased from 108.3 ± 32.4 μm to 30.6 ± 9.2 μm. The maximum cooling rate achieved during the laser cladding of the alloy was 44 × 10{sup 3} K s{sup −1}. The electron microscopy study of the alloy showed the presence of precipitates. The morphology of the disordered bcc (Fe, Cr)-rich precipitates in the ordered B2 (Al, Ni)-rich matrix changed in the dendritic and interdendritic regions from fine and spherical (with a diameter of less 100 nm) to spinodal (with the thickness being less than 100 nm). The LENS- produced AlCoCrFeNi alloy exhibited an average microhardness of approximately 543 HV0.5; this was approximately 13% higher than the hardness in the as-cast state and can be attributed to the grain refinemet in the LENS- produced alloy. Moreover, it was found that increasing the cooling rate during laser cladding increasess the microhardness of the alloy. - Highlights: • Laser-engineered net shaping is used to produce samples of AlCoCrFeNi alloy. • The alloy has a body-centred cubic (bcc)-derivative B2-ordered crystal structure. • Electron microscopy images of the alloy show the presence of precipitates. • The microhardness of the laser-clad alloy is higher than that of the as-cast alloy. • The cooling rate

  8. 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.

  9. Highlighting micrographic structures of uranium-zirconium alloys with 6 per cent of weight of Zr; Mise en evidence des structures micrographiques des alliages uranium-zirconium a 6 pour cent en poids de Zr

    Energy Technology Data Exchange (ETDEWEB)

    Bouleau, Maurice

    1961-01-17

    In order to study the transformation kinetics of U-Zr alloys with a Zr content of 6 per cent in weight, the authors searched for a slow enough electrolytic polishing bath, and for an attack and examination method to highlight martensite structures produced by austempering and water tempering, and ultra-fine decomposition structures obtained by austempering. The authors explain the choice of a perchloric-butyl glycol polishing bath, of an examination under polarized light or normal light after appropriate attacks. These studies are reported for annealed alloys, and for processed alloys with martensite or ultra-fine decomposition structures [French] L'etude de la cinetique de transformation des alliages U-Zr a 6 pc en poids de Zr a necessite la recherche d'un bain de polissage electrolytique assez lent et de methodes d'attaque et d'examen qui permettent la mise en evidence des structures martensitiques (provenant de trempes etagees ou de trempes a l'eau) et des structures de decomposition ultrafines (obtenues par trempes etagees). Nous nous sommes arretes dans notre choix: - sur un bain de polissage perchlorique-butyl glycol; sur des examens en lumiere polarisee ou en lumiere normale apres attaques appropriees (en cellule dans le meme electrolyte ou au tampon dans un bain phosphorique ethylene glycol). (auteur)

  10. 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.)

  11. State and perspectives of methodological support of materials research of products from Zirconium alloys for fuel rods and fuel assemblies of VVER

    International Nuclear Information System (INIS)

    Gusev, A.; Markelov, V.; Novikov, V.; Zheltkovskaya, T.; Malgin, A.; Shevyakov, A.; Bekrenev, S.

    2015-01-01

    The basic methodological framework for the study of the characteristics of zirconium products was created in JSC «VNIINM». The reliability of experiments confirmed the results of metrological certification procedures. Further development of methodological support of «VNIINM» for Zr products research is the development and validation of methods to determine: mechanical characteristics under internal pressure; Determination of Contractile Strain Ratio (CSR); Expansion Due to Compression (EDC); Plane Strain Tensile (PST); characteristics of resistance multi-cycle and low-cyclic fatigue; texture parameters using the orientation distribution function; the electrical characteristics of the oxide film by impedance

  12. Method of reducing zirconium

    International Nuclear Information System (INIS)

    Megy, J.A.

    1980-01-01

    A method was developed for making nuclear-grade zirconium from a zirconium compound, which ismore economical than previous methods since it uses aluminum as the reductant metal rather than the more expensive magnesium. A fused salt phase containing the zirconium compound to be reduced is first prepared. The fused salt phase is then contacted with a molten metal phase which contains aluminum and zinc. The reduction is effected by mutual displacment. Aluminum is transported from the molten metal phase to the fused salt phase, replacing zirconium in the salt. Zirconium is transported from the fused salt phase to the molten metal phase. The fused salt phase and the molten metal phase are then separated, and the solvent metal and zirconium are separated by distillation or other means. (DN)

  13. 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

  14. Preparation and characterisation of Co–Fe–Ni–M-Si–B (M = Zr, Ti) amorphous powders by wet mechanical alloying

    Energy Technology Data Exchange (ETDEWEB)

    Neamţu, B.V., E-mail: Bogdan.Neamtu@stm.utcluj.ro [Materials Science and Engineering Department, Technical University of Cluj-Napoca, 103-105, Muncii Avenue, 400641, Cluj-Napoca (Romania); Chicinaş, H.F.; Marinca, T.F. [Materials Science and Engineering Department, Technical University of Cluj-Napoca, 103-105, Muncii Avenue, 400641, Cluj-Napoca (Romania); Isnard, O. [Université Grenoble Alpes, Institut NEEL, F-38042, Grenoble (France); CNRS, Institut NEEL, 25 rue des martyrs, BP166, F-38042, Grenoble (France); Chicinaş, I. [Materials Science and Engineering Department, Technical University of Cluj-Napoca, 103-105, Muncii Avenue, 400641, Cluj-Napoca (Romania)

    2016-07-15

    Co-based amorphous alloys were prepared via wet mechanical alloying process starting from elemental powders. The reference alloy Co{sub 70}Fe{sub 4}Ni{sub 2}Si{sub 15}B{sub 9} (at. %) as well as the alloys derived from this composition by the substitution of 5 at.% of Zr or Ti for Si or B (Co{sub 70}Fe{sub 4}Ni{sub 2}Si{sub 15}B{sub 4}Zr{sub 5}, Co{sub 70}Fe{sub 4}Ni{sub 2}Si{sub 15}B{sub 4}Ti{sub 5}, Co{sub 70}Fe{sub 4}Ni{sub 2}Si{sub 10}B{sub 9}Zr{sub 5} and Co{sub 70}Fe{sub 4}Ni{sub 2}Si{sub 10}B{sub 9}Ti{sub 5}) are obtained in amorphous state, according to X-ray diffraction (XRD) investigation, after 40 h of milling. The calculated amount of amorphous fraction reaches 99% after 40 h of milling. The largest increase of the crystallisation temperature was induced by the substitution of Zr or Ti for Si while, regardless of the type of substitution, an important increase of the Curie temperature of the alloy was obtained. A Co-based solid solution, with Co{sub 2}Si and Co{sub 2}B phases, result after crystallisation of the amorphous alloys as proved by XRD investigations. Saturation magnetisation of the alloys decreases upon increasing milling time, however it remains larger than the saturation magnetisation of the reference alloy. This was discussed in correlation with the specificity of the wet mechanical alloying process and the influence of the chemical bonding between Co and metalloids atoms over the magnetic moment of Co. - Highlights: • Co–Fe–Ni–M-Si–B (M = Zr, Ti) amorphous powders were prepared by wet MA. • Amorphisation of the alloy is reached after 40 h of wet MA for any composition. • Magnetisation decrease upon increasing milling time. • Substituting 5% Zr/Ti for Si increases significantly the alloy's thermal stability. • Substitution of 5 at. % Zr/Ti for Si increases the saturation magnetisation by 20%.

  15. 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.)

  16. AE characterisation of damage in burn resistant β-Titanium alloy (Tl-25V-15CR-2AL-0.2C)

    International Nuclear Information System (INIS)

    Shukri Mohd; Bowen, P.

    2003-01-01

    Acoustic Emission (AE) has been used to monitor cleavage cracking in Beta Titanium alloys to gain a detailed understanding of damage mechanisms in such alloys. Damage has been characterised in situ by AE and fractography for fracture toughness and fatigue crack growth resistance tests. During fracture toughness testing, the formation of cleavage facets in as received and heat treated specimens has been monitored by using AE. The values of fracture toughness have also been compared. For as received specimens, loads were applied until fracture while for heat treated specimens, three different test sequences were employed. For the first test, the load was again applied until failure. This test was carried out to give the overall clip gauge displacement curve for this specimen and also to estimate the load at which 5% offset and any pop-in occurs. For the second test, the load was applied to the specimens and then stopped at the 5 % offset and for the last test the applied-load was removed just after pop-in. Both of these latter specimens were then post fatigued until failure and then the fractured surfaces of all specimens were analysed using SEM. For fatigue crack growth resistance tests, the tests were started at ΔK of 20 MPam 1/2 until failure. A test was also stopped at ΔK of 40 MPam 1/2 . During such tests, crack growth was monitored by measuring the potential drop across the crack. It is found that the AE technique allows the investigation of the onset of cleavage fracture in this Beta-Ti alloy. It is, in principle, possible to use this technique to monitor online the formation of such unstable cleavage cracking of this alloy. (Author)

  17. Design of multi materials combining crystalline and amorphous metallic alloys

    International Nuclear Information System (INIS)

    Volland, A.; Ragani, J.; Liu, Y.; Gravier, S.; Suéry, M.; Blandin, J.J.

    2012-01-01

    Highlights: ► Elaboration of multi materials associating metallic glasses and conventional crystalline alloys by co-deformation performed at temperatures close to the glass transition temperature of the metallic glasses. ► Elaboration of filamentary metal matrix composites with a core in metallic glass by co extrusion. ► Sandwich structures produced by co-pressing. ► Detection of atomic diffusion from the glass to the crystalline alloys during the processes. ► Good interfaces between the metallic glasses and the crystalline alloys, as confirmed by mechanical characterisation. - Abstract: Multi materials, associating zirconium based bulk metallic glasses and crystalline metallic alloys like magnesium alloys or copper are elaborated by co-deformation processing performed in the supercooled liquid regions (SLR) of the bulk metallic glasses. Two processes are investigated: co-extrusion and co-pressing. In the first case, filamentary composites with various designs can be produced whereas in the second case sandwich structures are obtained. The experimental window (temperature, time) in which processing can be carried out is directly related to the crystallisation resistance of the glass which requires getting information about the crystallisation conditions in the selected metallic glasses. Thermoforming windows are identified for the studied BMGs by thermal analysis and compression tests in their SLR. The mechanical properties of the produced multi materials are investigated thanks to specifically developed mechanical devices and the interfaces between the amorphous and the crystalline alloys are characterised.

  18. 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)

  19. Glow discharge preparation and electrooptical characterisation of amorphous silicon alloys for solar cells. Preparacion por descarga luminiscente y caracterizacion electrooptica de aleaciones de silicio amorfo para celulas solares

    Energy Technology Data Exchange (ETDEWEB)

    Carabe, J

    1990-11-01

    A study is presented, focused on the preparation and characterisation of hydrogenated amorphous silicon alloy thin films for their application as p type window layers in pin silicon solar cells. The preparation technique used was radio frequency glow discharge. The samples were characterised optically (visible, near infrared and infrared absorption spectrophotometry) and electrically (dark and photoconductivities at ambient temperature and as functions of temperature). The influence of each of the preparation parameters on film properties has been systematically studied. The results have been analysed according to the existing models. Chapter 1 is an introduction to the material in question and its photovoltaic applications. Chapter 2 describes the experimental procedure used. Capter 3 shows and discusses the most relevant results obtained in the study of intrinsic amorphous silicon, p type amorphous silicon and p type amorphous silicon carbide window layers, with special emphasis on the influence of the use of an alternative dopant gas: boron trifluoride. Finally, chapter 4 summarises the most relevant conclusions drawn from this research work. (Author)

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

    Energy Technology Data Exchange (ETDEWEB)

    Ribis, J

    2007-12-15

    The fuel rod cladding, strongly affected by microstructural changes due to irradiation such as high density of dislocation loops, is strained by the end-of-life fuel rod internal pressure and the potential release of fission gases and helium during dry storage. Within the temperature range that is expected during dry interim storage, cladding undergoes long term creep under over-pressure. So, in order to have a predictive approach of the behavior of zirconium alloys cladding in dry storage conditions it is essential to take into account: initial dislocation loops, thermal annealing of loops and creep straining due to over pressure. Specific experiments and modelling for irradiated samples have been developed to improve our knowledge in that field. A Zr-1%Nb-O alloy was studied using fine microstructural investigations and mechanical testing. The observations conducted by transmission electron microscopy show that the high density of loops disappears during a heat treatment. The loop size becomes higher and higher while their density falls. The microhardness tests reveal that the fall of loop density leads to the softening of the irradiated material. During a creep test, both temperature and applied stress are responsible of the disappearance of loops. The loops could be swept by the activation of the basal slip system while the prism slip system is inhibited. Once deprived of loops, the creep properties of the irradiated materials are closed to the non irradiated state, a result whose consequence is a sudden acceleration of the creep rate. Finally, a micro-mechanical modeling based on microscopic deformation mechanisms taking into account experimental dislocation loop analyses and creep test, was used for a predictive approach by constructing a deformation mechanism map of the creep behavior of the irradiated material. (author)

  1. Influence of heat treatment conditions on structure and corrosion fracture of welded joints of zirconium alloy with 2.5 % niobium in agressive media

    International Nuclear Information System (INIS)

    Goncharov, A.B.; Nerodenko, M.M.; Tkachenko, L.M.; Adeeva, L.I.

    1990-01-01

    Influence of heat treatment on corrosion resistance of Zr-2.5 % Nb alloy welded joints is studied. It is stated that alloy after annealing in β-region has maximum corrosion resistance in sulfuric acid. Corrosion resistance in acetic acid doesn't depend on heat treatment. The best operating characteristics in steam-water medium of high parameters have welded joints, structure and phase composition of which approach α-phase with fine-dispersed β Nb particles, uniformly distributed in grain matrix. Such structure is attained by annealing in α-region or quenching with the following annealing at 850 K

  2. Metallurgy of zirconium and hafnium

    International Nuclear Information System (INIS)

    Baryshnikov, N.V.; Geger, V.Eh.; Denisova, N.D.; Kazajn, A.A.; Kozhemyakin, V.A.; Nekhamkin, L.G.; Rodyakin, V.V.; Tsylov, Yu.A.

    1979-01-01

    Considered are those properties of zirconium and of hafnium, which are of practical interest for the manufacture of these elements. Systematized are the theoretical and the practical data on the procedures for thermal decomposition of zirconia and for obtaining zirconium dioxide and hafnium dioxide by a thermal decomposition of compounds and on the hydrometallurgical methods for extracting zirconium and hafnium. Zirconium and hafnium fluorides and chlorides production procedures are described. Considered are the iodide and the electrolytic methods of refining zirconium and hafnium

  3. Microstructural characterisation and constitutive behaviour of alloy RR1000 under fatigue and creep-fatigue loading conditions

    International Nuclear Information System (INIS)

    Stoecker, C.; Zimmermann, M.; Christ, H.-J.; Zhan, Z.-L.; Cornet, C.; Zhao, L.G.; Hardy, M.C.; Tong, J.

    2009-01-01

    Mechanical behaviour of a nickel-based superalloy, RR1000, has been investigated at 650 deg. C under cyclic and dwell loading conditions. The microstructural characteristics of the alloy have been studied using scanning electron microscopy (SEM) and transmission electron microscopy (TEM), and the distribution patterns of the dislocations and slip planes have been compared between samples tested under fatigue and creep-fatigue loading conditions. Constitutive behaviour of the alloy was described by a unified constitutive model, where both cyclic plastic and viscoplastic strains were represented by one inelastic strain. The results show that the precipitation state is very stable at 650 deg. C and only minor differences exist in the dislocation arrangements formed under pure fatigue and combined creep and fatigue conditions. Hence, a unified constitutive model seems to be justified in describing and predicting the constitutive behaviour in both cases.

  4. Quasi-static characterisation of trained pseudoelastic shape memory alloy wire subjected to cyclic loading: transformation kinetics

    DEFF Research Database (Denmark)

    Enemark, Søren; Santos, Ilmar

    2016-01-01

    A kinetic law for constitutive modelling of shape memory alloys is proposed in order to increase model predictability in comparison with experimental data. The proposed law is based on cubic Bézier curves and contains curvature controlling parameters. The kinetic law and also the Duhem–Madelung sub......-loop model are implemented in a state-of-the-art constitutive model based on the framework by Lagoudas and coworkers. The original and modified models are fitted to consistent experimental results from mechanical cyclic loading under isothermal conditions (0–800 MPa, 30–70 °C) of a trained pseudoelastic...... shape memory alloy wire. Quantitative measures of goodness of fit show that both models perform well, but use of the modified model results in 31% reduction of the residual standard deviation compared with the original model (21.4 versus 14.8 MPa) in model calibration and 23% in model validation...

  5. On the mechanical and electrical properties of copper-silver and copper-silver-zirconium alloys deposits manufactured by cold spray

    International Nuclear Information System (INIS)

    Coddet, Pierre; Verdy, Christophe; Coddet, Christian; Debray, François

    2016-01-01

    In this work, several copper alloy deposits were manufactured by cold spray with helium as accelerating and carrier gas. Electrical conductivity was measured to establish the potential of cold spray as a manufacturing process for high strength (>500 MPa) and high conductivity (>90% IACS) copper alloys. The deposits which are characterized by a low oxygen content (<200 ppm) and a low porosity level (<0.1%) present yield strength values up to about 700 MPa and electrical conductivity values up to 58.2 MS/m (100.3% IACS). Results show that, even if a compromise has to be made between the properties according to the objectives of the application, this additive manufacturing route appears suitable for the production of large copper alloys parts with high mechanical properties and high electrical and thermal conductivity. The role of alloy composition and post heat treatments on the strength and conductivity of the deposits was especially considered in this work. Cold spray deposits properties were finally compared with those obtained with other manufacturing routes.

  6. On the mechanical and electrical properties of copper-silver and copper-silver-zirconium alloys deposits manufactured by cold spray

    Energy Technology Data Exchange (ETDEWEB)

    Coddet, Pierre, E-mail: pierre-laurent.coddet@univ-orleans.fr [Laboratoire National des Champs Magnétiques Intenses (LNCMI – CNRS-UPS-INSA-UJF), 25 Rue des Martyrs, 38042 Grenoble (France); Verdy, Christophe; Coddet, Christian [UTBM, Site de Sévenans, 90010 Belfort Cedex (France); Debray, François [Laboratoire National des Champs Magnétiques Intenses (LNCMI – CNRS-UPS-INSA-UJF), 25 Rue des Martyrs, 38042 Grenoble (France)

    2016-04-26

    In this work, several copper alloy deposits were manufactured by cold spray with helium as accelerating and carrier gas. Electrical conductivity was measured to establish the potential of cold spray as a manufacturing process for high strength (>500 MPa) and high conductivity (>90% IACS) copper alloys. The deposits which are characterized by a low oxygen content (<200 ppm) and a low porosity level (<0.1%) present yield strength values up to about 700 MPa and electrical conductivity values up to 58.2 MS/m (100.3% IACS). Results show that, even if a compromise has to be made between the properties according to the objectives of the application, this additive manufacturing route appears suitable for the production of large copper alloys parts with high mechanical properties and high electrical and thermal conductivity. The role of alloy composition and post heat treatments on the strength and conductivity of the deposits was especially considered in this work. Cold spray deposits properties were finally compared with those obtained with other manufacturing routes.

  7. Preparation, microstructural evolution and properties of Ni–Zr intermetallic/Zr–Si ceramic reinforced composite coatings on zirconium alloy by laser cladding

    Energy Technology Data Exchange (ETDEWEB)

    Liu, Kun; Li, Yajiang, E-mail: yajli@sdu.edu.cn; Wang, Juan; Ma, Qunshuang; Li, Jishuai; Li, Xinyue

    2015-10-25

    NiZr{sub 2}–ZrSi–Zr{sub 5}(Si{sub x}Ni{sub 1−x}){sub 4}-ZrC intermetallic/ceramic reinforced composite coatings were in situ synthesized by laser cladding the pre-placed Ni–Cr–B–Si powder on zirconium substrate. Microstructure and phase constituents were investigated by X-ray diffraction (XRD), optical microscope (OM), scanning electron microscope (SEM) and energy dispersive spectroscopy (EDS). Microhardness tester and block-on-ring wear tester were employed to measure the hardness distribution and wear resistance of the intermetallic/ceramic reinforced composite coating. Results indicated that the multiphase of reinforcements includes Ni–Zr intermetallic compounds (e.g., NiZr and NiZr{sub 2}) and Zr–Si(C) ceramic phases (e.g., ZiSi, Zr{sub 5}Si{sub 4} and ZrC). Ni–Si clusters transforming to Zr–Si–Ni clusters at high temperature facilitated the forming of Zr{sub 5}(Si{sub x}Ni{sub 1−x}){sub 4} and during the growth of Zr{sub 5}(Si{sub x}Ni{sub 1−x}){sub 4}, the consumption of Zr atoms at the lateral interface of liquid/Zr{sub 5}(Si{sub x}Ni{sub 1−x}){sub 4} resulted into developing Zr-poor zone near Zr{sub 5}(Si{sub x}Ni{sub 1−x}){sub 4}. The microhardness and wear resistance of the coating were significantly improved by various reinforced phases in comparison to zirconium substrate. - Highlights: • NiZr{sub 2}–ZrSi–Zr{sub 5}(Si{sub x}Ni{sub 1−x}){sub 4}-ZrC compostie coating was in-situ synthesized. • Ni–Si clusters transforming resulted into developing Zr-poor zone near Zr{sub 5}(Si{sub x}Ni{sub 1−x}){sub 4}. • Reinforced phases significantly improve wear resistance of the coating.

  8. Characterisation of Oxides Formed on the Internal Surface of Steam Generator Tubes in Alloy 690 Corroded in the Primary Environment of Pressurised Water Reactors

    International Nuclear Information System (INIS)

    Carrette, Florence; Leclercq, Stephanie; Legras, Laurent

    2012-09-01

    Since the end of the 1990s, EDF R and D has been studying the phenomenon of corrosion product release from Steam Generator tubes in order to minimize the Source Term of the contamination and radiation exposure during operation and maintenance of Pressurised Water Reactors. With the BOREAL loop, release tests in primary water at 325 deg. C were performed on various Steam Generator tubes made of alloy 690. The experimental conditions of these tests (chemistry, temperature and hydraulics) were the same for all the tests but the results showed various behaviours towards release. For some tubes, the release was weak whereas for others, it was higher; the release rate of the tubes decreased more or less quickly with time. In order to explain these results, the internal surface of the tubes was characterised before and after the tests. Before the tests, various parameters were studied; the main parameters were the roughness, the impurities, the grain size and the cold work. The results demonstrated that it was not easy to quantify the influence of each parameter on release and to differentiate the tubes. A new parameter was proposed to characterise the internal extreme surface of SG tubes: the surface nano-hardness by nano-indentation measurements. The tubes were also observed and analysed by SEM, (X)TEM. Data obtained by (X)TEM revealed differences of the surface state (layer of perturbed microstructure, density of dislocations, grain size, impurities, initial oxide,...). After the tests, the oxides formed on the internal surface and the underlying material of the samples were characterised by SEM, (X)TEM and SIMS. The examinations showed various types of oxides. For some tubes, a duplex oxide scale was identified, for the others, only one oxide scale was observed. For equivalent durations of corrosion, the thickness of the enriched - chromium oxide layer can vary from 5 nm to 100 nm and the chemical composition can be different. The examinations of the underlying

  9. 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

  10. TiNi shape memory alloys: effects of the fabrication route, the oxygen content and the zirconium or hafnium additions on the metallurgical characteristics and the thermomechanical properties

    International Nuclear Information System (INIS)

    Olier, P.

    1996-01-01

    In order to promote the development of Ti-Ni shape memory alloys, we have studied the correlation between the fabrication route, the chemical composition (O 2 content, Zr or Hf additions), the metallurgical characteristics and the thermomechanical properties. A conventional sintering does not allow to obtain a homogeneous compound of pure Ti 50 Ni 50 alloy because of the occurrence of Kirkendall porosities which act as a diffusion barrier. An original process including combustion synthesis and hot-extrusion was successfully developed. Resulting products exhibit a smaller grain size (15-20μm) and an enhanced workability in comparison with products obtained by arc-melting and subsequent hot rolling. The presence of oxygen in equiatomic Ti-Ni alloy induces the oxide precipitation of Ti 4 Ni 2 O x type (with x ≤ 1). The precipitated particle fraction is proportional to the oxygen nominal content of the alloy. We show that the decrease of the transformation temperatures is correlated with the decrease of Ti in solid solution due to Ti 4 Ni 2 O precipitation. Moreover, we find that a fine and homogenous oxide dispersion is suitable to decrease the grain size during hot rolling and to enhance to the one way shape memory properties. An increase of the typical transformation temperatures is obtained through of Zr or Hf (in substitution to Ti). But, an increase of the hardness is measured, and consequently the workability of the ternary alloys becomes reduced. However, it is worthwhile to point out that a Ti 38 Ni 50 Hf 12 product obtained by arc melting and hot extrusion is able to fully recover an apparent plastic strain of more than 4% during tensile tests performed under special loading conditions. Such as behaviour is of great interest with respect to potential applications in a temperature range higher that 100 deg. C. (author)

  11. 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.

  12. SEPARATING HAFNIUM FROM ZIRCONIUM

    Science.gov (United States)

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

    1956-08-21

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

  13. 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

  14. Titanium and zirconium based wrought alloys and bulk metallic glasses for fluoride ion containing 11.5 M HNO3 medium

    International Nuclear Information System (INIS)

    Jayaraj, J.; Ningshen, S.; Mallika, C.; Kamachi Mudali, U.

    2016-01-01

    Aqueous reprocessing of plutonium-rich mixed oxide fuels require fluoride as a catalyst in boiling nitric acid for an effective dissolution of the spent fuel. The corrosion behavior of the candidate dissolver materials zircaloy-4 (Zr-4) and commercial pure titanium (CP-Ti grade 2) in boiling 11.5 M HNO 3 + 0.05 M NaF has been established. High corrosion rates were obtained for Zr- 4 and CP-Ti in nitric acid containing fluoride ions. Complexing the fluoride ions either with Al(NO 3 ) 3 or ZrO(NO 3 ) 2 aided in decreasing the corrosion rates of Zr-4 and CP-Ti. High corrosion resistance is claimed as one of the principal property of the amorphous alloy when compared to the crystalline alloy. Thus Ni 60 Nb 40 and Ni 60 Nb 30 Ta 10 amorphous ribbons were prepared and exposed in boiling 11.5 M HNO 3 and 11.5 M HNO 3 + 0.05 M NaF. In nitric acid these alloys did not show any sign of corrosion attack. XPS analysis confirmed that the passivity was due to the formation passive films of thickness ≈3 nm enriched with Nb 2 O 5 and of ≈1.5 nm enriched with both Nb 2 O 5 and Ta 2 O 5 on the respective surfaces of the ribbons. In boiling 11.5 M HNO 3 + 0.05 M NaF, severe corrosion attack was observed on Ni 60 Nb 40 ribbon, due to the instability of the oxide/metal interface. The Ni 60 Nb 30 Ta 10 amorphous ribbon exhibited corrosion resistance of at least an order of magnitude higher than that for Ni 60 Nb 40 ribbon

  15. Corrosion characterisation of laser beam and tungsten inert gas weldment of nickel base alloys: Micro-cell technique

    International Nuclear Information System (INIS)

    Abraham, Geogy J.; Kain, V.; Dey, G.K.; Raja, V.S.

    2015-01-01

    Highlights: • Grain matrix showed better corrosion resistance than grain boundary. • Microcell studies showed distinct corrosion behaviour of individual regions of weldment. • TIG welding resulted in increased stable anodic current density on weld fusion zone. • LB welding resulted in high stable anodic current density for heat affected zone. - Abstract: The electrochemical studies using micro-cell technique gave new understanding of electrochemical behaviour of nickel base alloys in solution annealed and welded conditions. The welding simulated regions depicted varied micro structural features. In case of tungsten inert gas (TIG) weldments, the weld fusion zone (WFZ) showed least corrosion resistance among all other regions. For laser beam (LB) weldments it was the heat-affected zone (HAZ) that showed comparatively high stable anodic current density. The high heat input of TIG welding resulted in slower heat dissipation hence increased carbide precipitation and segregation in WFZ resulting in high stable anodic current density

  16. Effect of hydrogen and oxygen content on the embrittlement of Zr alloys

    International Nuclear Information System (INIS)

    Griger, A.; Hozer, Z.; Matus, L.; Vasaros, L.; Horvath, M.

    2001-01-01

    An experimental study is carried out in the KFKI Atomic Energy Research Institute in order to clear up the role of oxidation and hydrogen uptake in the embrittlement process. Russian E110 type Zr1%Nb and Zircaloy-4 claddings are used as test materials. The differences between the properties of two alloys are examined. The sample preparation covered the following cases: oxidation in Ar+O 2 atmosphere; hydrogen uptake of as received and pre-oxidised samples (in Ar+O 2 atmosphere); oxidation in steam. The oxidation in Ar+O 2 and the subsequent hydrogen uptake procedure make possible the production of samples with well-characterized hydrogen and oxygen content. Corrosion treated ring samples of 8 mm height are examined in ring compression tests. The force-deformation curves are recorded and the crushing force and deformation are determined. The relative deformation is used for the characterisation of embrittlement level. The results of experiments provide detailed information about the effect of hydrogen and oxygen content on the embrittlement of zirconium alloys. The conclusions are: 1) hydrogen seems to play a more important role in the embrittlement of zirconium alloys than oxygen; 2) the Zircaloy-4 alloy becomes brittle at lower hydrogen content than the Zr1%Nb; 3) under steam oxidation conditions the Zr1%Nb alloy takes up much more hydrogen and becomes more brittle than the Zircaloy-4

  17. Effects of zirconium and nitrogen plasma immersion ion implantation on the electrochemical corrosion behavior of Mg–Y–RE alloy in simulated body fluid and cell culture medium

    International Nuclear Information System (INIS)

    Jamesh, Mohammed Ibrahim; Wu, Guosong; Zhao, Ying; Jin, Weihong; McKenzie, David R.; Bilek, Marcela M.M.; Chu, Paul K.

    2014-01-01

    Highlights: • Dual Zr and N plasma ion implantation are conducted on WE43Mg alloy. • Zr and N implanted WE43 (ZrN-WE43) enhanced corrosion resistance in cell culture medium. • ZrN-WE43 enhanced corrosion resistance in simulated body fluid (SBF). • ZrN-WE43 shows near capacitive impedance spectra in cell culture medium. • Calcium phosphate is formed on the corrosion product. - Abstract: The effects of dual Zr and N plasma immersion ion implantation (PIII) on the corrosion behavior of WE43Mg alloy are evaluated in simulated body fluid (SBF) and cell culture medium (cDMEM). Zr and N PIII improves the corrosion resistance of WE43 which exhibits smaller i corr , larger R 1 and R 2 , smaller CPE 2 , and larger phase angle maxima in SBF and cDMEM. The Zr and N PIII WE43 samples exhibit 12-folds decrease in i corr in SBF and 71-folds decrease in i corr with near capacitive EIS in cDMEM. Analysis of the corrosion products reveals calcium phosphate

  18. Magnetic characterisation of large grain, bulk Y–Ba–Cu–O superconductor–soft ferromagnetic alloy hybrid structures

    International Nuclear Information System (INIS)

    Philippe, M.P.; Fagnard, J.-F.; Kirsch, S.; Xu, Z.; Dennis, A.R.; Shi, Y.-H.; Cardwell, D.A.; Vanderheyden, B.; Vanderbemden, P.

    2014-01-01

    Highlights: • Large grain, bulk YBaCuO superconductor (SC) combined with ferromagnetic elements. • The flux lines curve outwards through the ferromagnet in the remanent state. • The trapped field in the SC is enhanced by the presence of the ferromagnet. • The effects of the SC and the ferromagnet add when the ferromagnet is saturated. - Abstract: Large grain, bulk Y–Ba–Cu–O (YBCO) high temperature superconductors (HTS) have significant potential for use in a variety of practical applications that incorporate powerful quasi-permanent magnets. In the present work, we investigate how the trapped field of such magnets can be improved by combining bulk YBCO with a soft FeNi, ferromagnetic alloy. This involves machining the alloy into components of various shapes, such as cylinders and rings, which are attached subsequently to the top surface of a solid, bulk HTS cylinder. The effect of these modifications on the magnetic hysteresis curve and trapped field of the bulk superconductor at 77 K are then studied using pick-up coil and Hall probe measurements. The experimental data are compared to finite element modelling of the magnetic flux distribution using Campbell’s algorithm. Initially we establish the validity of the technique involving pick-up coils wrapped around the bulk superconductor to obtain its magnetic hysteresis curve in a non-destructive way and highlight the difference between the measured signal and the true magnetization of the sample. We then consider the properties of hybrid ferromagnet/superconductor (F/S) structures. Hall probe measurements, together with the results of the model, establish that flux lines curve outwards through the ferromagnet, which acts, effectively, like a magnetic short circuit. Magnetic hysteresis curves show that the effects of the superconductor and the ferromagnet simply add when the ferromagnet is saturated fully by the applied field. The trapped field of the hybrid structure is always larger than that of the

  19. Zirconium isotope separation process

    International Nuclear Information System (INIS)

    Peterson, S.H.; Lahoda, E.J.

    1988-01-01

    A process is described for reducing the amount of zirconium 91 isotope in zirconium comprising: forming a first solution of (a) a first solvent, (b) a scavenger, and (c) a zirconium compound which is soluble in the first solvent and reacts with the scavenger when exposed to light of a wavelength of 220 to 600 nm; irradiating the first solution with light at the wavelength for a time sufficient to photoreact a disproportionate amount of the zirconium compound containing the zirconium 91 isotope with the scavenger to form a reaction product in the first solution; contacting the first solution, while effecting the irradiation, with a second solvent which is immiscible with the first solvent, which the second solvent is a preferential solvent for the reaction product relative to the first solvent, such that at least a portion of the reaction product is transferred to the second solvent to form a second solution; and separating the second solution from the first solution after the contacting

  20. Long term corrosion of iron and non alloy or low alloy steels in clay soils. Physico-chemical characterisation and electrochemical study of archaeological analogues

    International Nuclear Information System (INIS)

    Pons, Emmanuelle

    2002-01-01

    Archaeological objects of Gallo-Roman and Merovingian time, and from a battlefield of World War 1, were studied to better understand long term corrosion phenomena of iron in clay soils. This study is part of the French national program about nuclear waste deep repository, conducted by the ANDRA (French national Agency for Radioactive Waste Management). Iron archaeological analogues make a valuable contribution to the specifying of containers for long lived and high level wastes (HLWs), because they provide access to the considered time scale. The experimental issue is divided into two major parts: - a physico-chemical characterisation of corrosion products, by Raman spectroscopy; - an electrochemical study of the behaviour of the different corrosion layers. Although the metallic material is different between ancient artefacts (ferrite) and 1914-1918 remains (hypo-eutectoid steels), the same stable phases are identified in their corrosion products: mainly iron oxides and oxi-hydroxides. From a macroscopic point of view, these products are staggered into two layers: an internal one, and an external one, which contains soil markers. Under the microscope, a complex composite structure appears. Goethite a-FeOOH, which was identified on each object, is frequently in contact with the metal core. The average corrosion rate in the burial environment, deduced from the layers thickness, highlights a significant slowdown of corrosion after the first burial time, about one century. The electrochemical study showed the predominant role of transport phenomena in the pores of corrosion layers. The behaviour of the metal - internal layer system is well explained by a model of porous electrode (De Levie theory). Despite its porosity, the internal layer is protective, as it leads to a significant decrease of the corrosion rate (about ten time). (author) [fr

  1. 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

  2. Purification of zirconium concentrates

    International Nuclear Information System (INIS)

    Brown, A.E.P.

    1976-01-01

    A commercial grade ZrO 2 and an ammonium uranate (yellow cake) are obtained from the caldasito ore processing. This ore is found in the Pocos de Caldas Plateau, State of Minas Gerais, Brazil. Caldasito is an uranigerous zirconium ore, a mixture of zircon and baddeleyite and contains 60% ZrO 2 and 0,3% U 3 O 8 . The chemical opening of the ore was made by alkaline fusion with NaOH at controlled temperature. The zirconium-uranium separation took place by a continuous liquid-liquid extraction in TBP-varsol-HNO 3 -H 2 O system. The raffinate containing zirconium + impurities (aluminium, iron and titanium) was purified by an ion exchange operation using a strong cationic resin [pt

  3. 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)

  4. 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.)

  5. 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

  6. Irradiation induced effects in zirconium (A review)

    International Nuclear Information System (INIS)

    Madden, P.K.

    1975-06-01

    Irradiation creep in zirconium and its alloys is comprehensively discussed. The main theories are outlined and the gaps between them and the observed creep behaviour, indicated. Although irradiation induced point defects play an important role, effects due to irradiation induced dislocation loops seem insignificant. The experimental results suggest that microstructural variations due to prior cold-working or hydrogen injection perturb the irradiation growth and the irradiation creep of zircaloy. Further investigations into these areas are required. One disadvantage of creep experiments lies in their duration. The possibility of accelerated experiments using ion implantation or electron irradiation is examined in the final section, and its possible advantages and disadvantages are outlined. (author)

  7. The Effect of Peak Temperatures and Hoop Stresses on Hydride Reorientations of Zirconium Alloy Cladding Tubes under Interim Dry Storage Condition

    International Nuclear Information System (INIS)

    Cha, Hyun Jin; Jang, Ki Nam; Kim, Kyu Tae

    2016-01-01

    In this study, the effect of peak temperatures and hoop tensile stresses on hydride reorientation in cladding was investigated. It was shown that the 250ppm-H specimens generated larger radial hydride fractions and longer radial hydrides than the 500ppm-H ones. The precipitated hydride in radial direction severely degrades mechanical properties of spent fuel rod. Hydride reorientation is related to cladding material, cladding temperature, hydrogen contents, thermal cycling, hoop stress and cooling rate. US NRC established the regulation on cladding temperature during the dry storage, which is the maximum fuel cladding temperature should not exceed 400 .deg. C for all fuel burnups under normal conditions of storage. However, if it is proved that the best estimate cladding hoop stress is equal to or less than 90MPa for the temperature limit proposed, a higher short-term temperature limit is allowed for low burnup fuel. In this study, 250ppm and 500ppm hydrogen-charged Zr-Nb alloy cladding tubes were selected to evaluate the effect of peak temperatures and hoop tensile stresses on the hydride reorientation during the dry storage. In order to evaluate threshold stresses in relation to various peak temperatures, four peak temperatures of 250, 300, 350, and 400 .deg. C and three tensile hoop stresses of 80, 100, 120MPa were selected.

  8. 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

  9. 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

  10. Study of the microstructural and mechanical properties of titanium-niobium-zirconium based alloys processed with hydrogen and powder metallurgy for use in dental implants; Estudo das propriedades mecanicas e microestruturais de ligas a base de titanio-niobiozirconio processados com hidrogenio e metalurgia do po para utilizacao em implantes dentarios

    Energy Technology Data Exchange (ETDEWEB)

    Duvaizem, Jose Helio

    2009-07-01

    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)

  11. 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.

  12. Finite element modelling of shot peening and peen forming processes and characterisation of peened AA2024-T351 aluminium alloy

    Science.gov (United States)

    Gariepy, Alexandre

    The main purpose of this thesis was to develop and validate finite element (FE) simulation tools for shot peening and peen forming. The specific aim was to achieve quantitatively accurate predictions for both processes and demonstrate the potential of reliable FE modelling for scientific investigation and industrial applications. First, an improved dynamic impact model that takes into account the stochastic nature of shot peening was proposed by carefully studying its dimensions, introducing a dispersion of shot sizes and significantly reducing its computational cost. In addition, cyclic mechanical testing was conducted to define a suitable material constitutive theory for aluminium alloy (AA) 2024-T3/T351 subjected to shot peening. By combining a realistic shot peening model with an appropriate material law, fairly good residual stress predictions were achieved for three different sets of shot peening parameters. Second, an experimental and numerical characterization of AA2024-T351 shot peened with parameters representative of fatigue life improvement applications was conducted. Multiple techniques, such as micro-indentation, residual stress determination and electron backscatter diffraction, were combined to gain a better understanding of the influence of shot peening on the material. The potential uses of finite element simulation to complement experimental data were also studied. The material heterogeneity arising from the random impact sequence was investigated and it was found that the impact modelling methodology could provide useful information on such heterogeneities. Third, a novel peen forming simulation methodology was introduced. The impact model provided the necessary input data as part of a multiscale approach. Numerically calculated unbalanced induced stress profiles were input into shell elements and the deformed shape after peen forming was computed as a springback analysis. In addition, a simple interpolation method was proposed to model the

  13. Solvent extraction of zirconium

    International Nuclear Information System (INIS)

    Kim, S.S.; Yoon, J.H.

    1981-01-01

    The extraction of zirconium(VI) from an aqueous solution of constant ionic strength with versatic acid-10 dissolved in benzen was studied as a function of pH and the concentration of zirconium(VI) and organic acid. The effects of sulphate and chlorine ions on the extraction of the zirconium(VI) were briefly examined. It was revealed that (ZrOR 2 .2RH) is the predominant species of extracted zirconium(VI) in the versatic acid-10. The chemical equation and the apparent equilibrium constants thereof have been determined as follows. (ZrOsup(2+))aq+ 2(R 2 H 2 )sub(org) = (ZrOR 2 .2RH)sub(org)+2(H + )aq Ksub(Zr) = (ZrOR 2 .2RH)sub(org)(H + ) 2 /(ZrOsup(2+))sub(aq)(R 2 H 2 )sup(2)sub(org) = 3.3 x 10 -7 . The synergistic effects of TBP and D2EHPA were also studied. In the mixed solvent with 0.1M TBP, the synergistic effect was observed, while the mixed solvent with D2EHPA showed the antisynergistic effect. (Author)

  14. 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%.

  15. 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)

  16. 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)

  17. 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.

  18. 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.

  19. 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

  20. Plasma arc melting of zirconium

    International Nuclear Information System (INIS)

    Tubesing, P.K.; Korzekwa, D.R.; Dunn, P.S.

    1997-01-01

    Zirconium, like some other refractory metals, has an undesirable sensitivity to interstitials such as oxygen. Traditionally, zirconium is processed by electron beam melting to maintain minimum interstitial contamination. Electron beam melted zirconium, however, does not respond positively to mechanical processing due to its large grain size. The authors undertook a study to determine if plasma arc melting (PAM) technology could be utilized to maintain low interstitial concentrations and improve the response of zirconium to subsequent mechanical processing. The PAM process enabled them to control and maintain low interstitial levels of oxygen and carbon, produce a more favorable grain structure, and with supplementary off-gassing, improve the response to mechanical forming

  1. Zirconium microstructures: uncharted possibilities

    International Nuclear Information System (INIS)

    Samajdar, I.; Kumar, Gulshan; Singh, Jaiveer; Lodh, Arijit; Srivastava, D.; Tewari, R.; Dey, G.K.; Saibaba, N.

    2015-01-01

    The 'conventional' Zirconium microstructures can be significantly extended with information on: (i) microtexture, (ii) residual stresses and (iii) local mechanical properties. Though these involve different tools, but a consolidated microstructure can be crated. This is the theme of this presentation. Examples of this consolidated picture will be made from deformation twinning, recovery-recrystallization, burst ductility and orientation versus solid solution hardening. (author)

  2. Zirconium elasticity modules

    International Nuclear Information System (INIS)

    Vavra, G.

    1978-01-01

    Considered are the limit and the intermediate values of the Young modulus E, modulus of shear G and of linear modulus of compression K obtainable at various temperatures (4.2 to 1133 K) for single crystals of α-zirconium. Determined and presented are the corrected isotropic elasticity characteristics of E, G, K over the above range of temperatures of textured and non-textured α-Zr

  3. Beryllium and zirconium

    International Nuclear Information System (INIS)

    Salesse, Marc

    1959-01-01

    Pure beryllium and zirconium, both isolated at about the same date but more than a century ago remained practically unused for eighty years. Fifteen years ago they were released from this state of inactivity by atomic energy, which made them into current metal a with an annual production which runs into tens of tons for the one and thousands for the other. The reasons for this promotion promise well for the future of the two metals, which moreover will probably find additional uses in other branches of industry. The attraction of beryllium and zirconium for atomic energy is easily explained. The curve of figure 1 gives the price per gram of uranium-235 as a function of enrichment: this price increases by about a factor of 3 on passing from natural uranium (0, 7 percent 235 U) to almost pure uranium-235. Because of their tow capture cross-section beryllium and zirconium make it possible, or at least easier, to use natural uranium and they thus enjoy an advantage the extent of which must be calculated for each reactor or fuel element project, but which is generally considerable. It will be seen later that this advantage should be based on figures which are even more favourable that would appear from the simple ratio 3 of the price of pure uranium- 235 contained in natural uranium. Reprint of a paper published in 'Industries Atomiques' - n. 1-2, 1959

  4. Chemical and electrical characterisation of the segregation of Al from a CuAl alloy (90%:10% wt) with thermal anneal

    Energy Technology Data Exchange (ETDEWEB)

    Byrne, C., E-mail: conor.byrne2@mail.dcu.ie [School of Physical Sciences, Dublin City University, Dublin 9 (Ireland); Brady, A.; Walsh, L.; McCoy, A.P.; Bogan, J. [School of Physical Sciences, Dublin City University, Dublin 9 (Ireland); McGlynn, E. [School of Physical Sciences, National Centre for Plasma Science and Technology, Dublin City University, Dublin 9 (Ireland); Rajani, K.V. [School of Electronic Engineering, Dublin City University, Dublin 9 (Ireland); Hughes, G. [School of Physical Sciences, Dublin City University, Dublin 9 (Ireland); School of Physical Sciences, National Centre for Plasma Science and Technology, Dublin City University, Dublin 9 (Ireland)

    2016-01-29

    A copper–aluminium (CuAl) alloy (90%:10% wt) has been investigated in relation to segregation of the alloying element Al, from the alloy bulk during vacuum anneal treatments. X-ray photoelectron spectroscopy (XPS) measurements were used to track the surface enrichment of Al segregating from the alloy bulk during in situ ultra-high vacuum anneals. Secondary ion mass spectroscopy (SIMS) indicates a build-up of Al at the surface of the annealed alloy relative to the bulk composition. Metal oxide semiconductor (MOS) CuAl/SiO{sub 2}/Si structures show a shift in flatband voltage upon thermal anneal consistent with the segregation of the Al to the alloy/SiO{sub 2} interface. Electrical four point probe measurements indicate that the segregation of Al from the alloy bulk following thermal annealing results in a decrease in film resistivity. X-ray diffraction data shows evidence for significant changes in crystal structure upon annealing, providing further evidence for expulsion of Al from the alloy bulk. - Highlights: • CuAl alloy (90%:Al 10% wt) deposited and vacuum annealed • XPS and SIMS data show segregation of Al from the alloy bulk. • Chemical changes seen indicate the reduction of Cu oxide and growth of Al Oxide. • Electrical measurements indicate a chemical change at the metal/SiO{sub 2} interface. • All data consistent with Cu diffusion barrier layer formed.

  5. 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.

  6. Process for purifying zirconium sponge

    International Nuclear Information System (INIS)

    Abodishish, H.A.M.; Kimball, L.S.

    1992-01-01

    This patent describes a Kroll reduction process wherein a zirconium sponge contaminated with unreacted magnesium and by-product magnesium chloride is produced as a regulus, a process for purifying the zirconium sponge. It comprises: distilling magnesium and magnesium chloride from: a regulus containing a zirconium sponge and magnesium and magnesium chloride at a temperature above about 800 degrees C and at an absolute pressure less than about 10 mmHg in a distillation vessel to purify the zirconium sponge; condensing the magnesium and the magnesium chloride distilled from the zirconium sponge in a condenser; and then backfilling the vessel containing the zirconium sponge and the condenser containing the magnesium and the magnesium chloride with a gas; recirculating the gas between the vessel and the condenser to cool the zirconium sponge from above about 800 degrees C to below about 300 degrees C; and cooling the recirculating gas in the condenser containing the condensed magnesium and the condensed magnesium chloride as the gas cools the zirconium sponge to below about 300 degrees C

  7. Oxygen stabilized zirconium-vanadium-iron alloy

    International Nuclear Information System (INIS)

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

    1982-01-01

    An oxygen stabilized intermetallic compound having the formula (Zrsub(1-x)Tisub(x))sub(2-u)(Vsub(1-y)Fesub(y))Osub(z) where x = 0.0 to 0.9, y = 0.01 to 0.9, z = 0.25 to 0.5 and u = 0 to 1. The compound is capable of reversibly sorbing hydrogen at temperatures from -196 deg C to 200 deg C at pressures down to 10 - 6 torr. The compound is suitable for use as a hydrogen getter in low pressure, high temperature applications such as magnetic confinement fusion devices, and the iron content may be substituted by nickel, cobalt or manganese. (author)

  8. Oxygen stabilized zirconium-vanadium-iron alloy

    International Nuclear Information System (INIS)

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

    1982-01-01

    An oxygen stabilized intermetallic compound having the formula (Zrsub(1-x)Tisub(x))sub(2-u)(Vsub(1-y)Fesub(y))Osub(z) where x=0.0 to 0.9, y=0.01 to 0.9, z=0.25 to 0.5 and u=0 to 1. The compound is capable of reversibly sorbing hydrogen at temperatures from -196 0 C to 200 0 C at pressures down to 10 - 6 torr. The compound is suitable for use as a hydrogen getter in low pressure, high temperature applications such as magnetic confinement fusion devices, and the iron content may be substituted by nickel, cobalt or manganese. (author)

  9. ZIRCONIUM PHOSPHATE ADSORPTION METHOD

    Science.gov (United States)

    Russell, E.R.; Adamson, A.S.; Schubert, J.; Boyd, G.E.

    1958-11-01

    A method is presented for separating plutonium values from fission product values in aqueous acidic solution. This is accomplished by flowing the solutlon containing such values through a bed of zirconium orthophosphate. Any fission products adsorbed can subsequently be eluted by washing the column with a solution of 2N HNO/sub 3/ and O.lN H/sub 3/PO/sub 4/. Plutonium values may subsequently be desorbed by contacting the column with a solution of 7N HNO/sub 3/ .

  10. Modification in band gap of zirconium complexes

    Energy Technology Data Exchange (ETDEWEB)

    Sharma, Mayank, E-mail: mayank30134@gmail.com; Singh, J.; Chouhan, S. [Department of Physics, ISLE, IPS Academy, Indore (M.P.) (India); Mishra, A. [School of Physics, Devi Ahilya Vishwavidyalaya, Indore (M.P.) (India); Shrivastava, B. D. [Govt. P. G. College, Biora (M.P.) (India)

    2016-05-06

    The optical properties of zirconium complexes with amino acid based Schiff bases are reported here. The zirconium complexes show interesting stereo chemical features, which are applicable in organometallic and organic synthesis as well as in catalysis. The band gaps of both Schiff bases and zirconium complexes were obtained by UV-Visible spectroscopy. It was found that the band gap of zirconium complexes has been modified after adding zirconium compound to the Schiff bases.

  11. 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)

  12. 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

  13. 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)

  14. 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...

  15. Measurements of the effective total and resonance absorption cross sections for zircaloy-2 and zirconium

    Energy Technology Data Exchange (ETDEWEB)

    Kocic, A; Markovic, V [Institute of nuclear sciences Boris Kidric, Vinca, Beograd (Yugoslavia)

    1967-04-15

    Zirconium and zircaloy-2 alloy, as constructive materials, have found wide