WorldWideScience

Sample records for uranium alloy specimens

  1. Machining of uranium and uranium alloys

    International Nuclear Information System (INIS)

    Morris, T.O.

    1981-01-01

    Uranium and uranium alloys can be readily machined by conventional methods in the standard machine shop when proper safety and operating techniques are used. Material properties that affect machining processes and recommended machining parameters are discussed. Safety procedures and precautions necessary in machining uranium and uranium alloys are also covered. 30 figures

  2. Irradiation Stability of Uranium Alloys at High Exposures

    International Nuclear Information System (INIS)

    McDonell, W.R.

    2001-01-01

    Postirradiation examinations were begun of a series of unrestrained dilute uranium alloy specimens irradiated to exposures up to 13,000 MWD/T in NaK-containing stainless steel capsules. This test, part of a program of development of uranium metal fuels for desalination and power reactors sponsored by the Division of Reactor Development and Technology, has the objective of defining the temperature and exposure limits of swelling resistance of the alloyed uranium. This paper discusses those test results

  3. Corrosion testing of uranium silicide fuel specimens

    International Nuclear Information System (INIS)

    Bourns, W.T.

    1968-09-01

    U 3 Si is the most promising high density natural uranium fuel for water-cooled power reactors. Power reactors fuelled with this material are expected to produce cheaper electricity than those fuelled with uranium dioxide. Corrosion tests in 300 o C water preceded extensive in-reactor performance tests of fuel elements and bundles. Proper heat-treatment of U-3.9 wt% Si gives a U 3 5i specimen which corrodes at less than 2 mg/cm 2 h in 300 o C water. This is an order of magnitude lower than the maximum corrosion rate tolerable in a water-cooled reactor. U 3 Si in a defected unbonded Zircaloy-2 sheath showed only a slow uniform sheath expansion in 300 o C water. All tests were done under isothermal conditions in an out-reactor loop. (author)

  4. Corrosion testing of uranium silicide fuel specimens

    Energy Technology Data Exchange (ETDEWEB)

    Bourns, W T

    1968-09-15

    U{sub 3}Si is the most promising high density natural uranium fuel for water-cooled power reactors. Power reactors fuelled with this material are expected to produce cheaper electricity than those fuelled with uranium dioxide. Corrosion tests in 300{sup o}C water preceded extensive in-reactor performance tests of fuel elements and bundles. Proper heat-treatment of U-3.9 wt% Si gives a U{sub 3}5i specimen which corrodes at less than 2 mg/cm{sup 2} h in 300{sup o}C water. This is an order of magnitude lower than the maximum corrosion rate tolerable in a water-cooled reactor. U{sub 3}Si in a defected unbonded Zircaloy-2 sheath showed only a slow uniform sheath expansion in 300{sup o}C water. All tests were done under isothermal conditions in an out-reactor loop. (author)

  5. Oxidation of uranium and uranium alloys

    International Nuclear Information System (INIS)

    Orman, S.

    1976-01-01

    The corrosion behaviour of uranium in oxygen, water and water + oxygen mixtures is compared and contrasted. A considerable amount of work, much of it conflicting, has been published on the U + H 2 O and U + H 2 O + O 2 systems. An attempt has been made to summarise this data and to explain the reasons for the lack of agreement between the experimental results. The evidence for the mechanism involving OH - ion diffusion as the reacting entity in both the U + H 2 O and U + O 2 + H 2 O reactions is advanced. The more limited corrosion data on some lean uranium alloys and on some higher addition alloys referred to as stainless materials is summarised together with some previously unreported results obtained with these materials at AWRE. The data indicates that in the absence of oxygen the lean alloys behave in a similar manner to uranium and evolve hydrogen in approximately theoretical quantities. But the stainless alloys absorb most of the product hydrogen and assessments of reactivity based on hydrogen evolution would be very inaccurate. The direction that future corrosion work on these materials should take is recommended

  6. Texture in low-alloyed uranium alloys

    International Nuclear Information System (INIS)

    Sariel, J.

    1982-08-01

    The dependence of the preferred orientation of cast and heat-treated polycrystalline adjusted uranium and uranium -0.1 w/o chromium alloys on the production process was studied. The importance of obtaining material free of preferred orientation is explained, and a survey of the regular methods to determine preferred orientation is given. Dilatometry, tensile testing and x-ray diffraction were used to determine the extent of the directionality of these alloys. Data processing showed that these methods are insufficient in a case of a material without any plastic forming, because of unreproducibility of results. Two parameters are defined from the results of Schlz's method diffraction test. These parameters are shown theoretically and experimentally (by extreme-case samples) to give the deviation from isotropy. Application of these parameters to the examined samples showes that cast material has preferred orientation, though it is not systematic. This preferred orientation was reduced by adequate heat treatments

  7. Process for continuous production of metallic uranium and uranium alloys

    Science.gov (United States)

    Hayden, Jr., Howard W.; Horton, James A.; Elliott, Guy R. B.

    1995-01-01

    A method is described for forming metallic uranium, or a uranium alloy, from uranium oxide in a manner which substantially eliminates the formation of uranium-containing wastes. A source of uranium dioxide is first provided, for example, by reducing uranium trioxide (UO.sub.3), or any other substantially stable uranium oxide, to form the uranium dioxide (UO.sub.2). This uranium dioxide is then chlorinated to form uranium tetrachloride (UCl.sub.4), and the uranium tetrachloride is then reduced to metallic uranium by reacting the uranium chloride with a metal which will form the chloride of the metal. This last step may be carried out in the presence of another metal capable of forming one or more alloys with metallic uranium to thereby lower the melting point of the reduced uranium product. The metal chloride formed during the uranium tetrachloride reduction step may then be reduced in an electrolysis cell to recover and recycle the metal back to the uranium tetrachloride reduction operation and the chlorine gas back to the uranium dioxide chlorination operation.

  8. Process for continuous production of metallic uranium and uranium alloys

    Science.gov (United States)

    Hayden, H.W. Jr.; Horton, J.A.; Elliott, G.R.B.

    1995-06-06

    A method is described for forming metallic uranium, or a uranium alloy, from uranium oxide in a manner which substantially eliminates the formation of uranium-containing wastes. A source of uranium dioxide is first provided, for example, by reducing uranium trioxide (UO{sub 3}), or any other substantially stable uranium oxide, to form the uranium dioxide (UO{sub 2}). This uranium dioxide is then chlorinated to form uranium tetrachloride (UCl{sub 4}), and the uranium tetrachloride is then reduced to metallic uranium by reacting the uranium chloride with a metal which will form the chloride of the metal. This last step may be carried out in the presence of another metal capable of forming one or more alloys with metallic uranium to thereby lower the melting point of the reduced uranium product. The metal chloride formed during the uranium tetrachloride reduction step may then be reduced in an electrolysis cell to recover and recycle the metal back to the uranium tetrachloride reduction operation and the chlorine gas back to the uranium dioxide chlorination operation. 4 figs.

  9. Fuel powder production from ductile uranium alloys

    International Nuclear Information System (INIS)

    Clark, C.R.; Meyer, M.K.

    1998-01-01

    Metallic uranium alloys are candidate materials for use as the fuel phase in very-high-density LEU dispersion fuels. These ductile alloys cannot be converted to powder form by the processes routinely used for oxides or intermetallics. Three methods of powder production from uranium alloys have been investigated within the US-RERTR program. These processes are grinding, cryogenic milling, and hydride-dehydride. In addition, a gas atomization process was investigated using gold as a surrogate for uranium. (author)

  10. 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.

  11. Fracture characteristics of uranium alloys by scanning electron microscopy

    International Nuclear Information System (INIS)

    Koger, J.W.; Bennett, R.K. Jr.

    1976-10-01

    The fracture characteristics of uranium alloys were determined by scanning electron microscopy. The fracture mode of stress-corrosion cracking (SCC) of uranium-7.5 weight percent niobium-2.5 weight percent zirconium (Mulberry) alloy, uranium--niobium alloys, and uranium--molybdenum alloys in aqueous chloride solutions is intergranular. The SCC fracture surface of the Mulberry alloy is characterized by very clean and smooth grain facets. The tensile-overload fracture surfaces of these alloys are characteristically ductile dimple. Hydrogen-embrittlement failures of the uranium alloys are brittle and the fracture mode is transgranular. Fracture surfaces of the uranium-0.75 weight percent titanium alloys are quasi cleavage

  12. Corrosion resistant coatings for uranium and uranium alloys

    International Nuclear Information System (INIS)

    Weirick, L.J.; Lynch, C.T.

    1977-01-01

    Coatings to prevent the corrosion of uranium and uranium alloys are considered in two military applications: kinetic energy penetrators and aircraft counterweights. This study, which evaluated organic films and metallic coatings, demonstrated that the two most promising coatings are based on an electrodeposited nickel system and a galvanized zinc system

  13. Internal hydrogen embrittlement of gamma-stabilized uranium alloys

    International Nuclear Information System (INIS)

    Powell, G.L.; Koger, J.W.; Bennett, R.K.; Williamson, A.L.; Hemperly, V.C.

    1976-01-01

    Relationships between the tensile ductility and fracture characteristics of as-quenched, gamma-stabilized uranium alloys (uranium--10 wt percent molybdenum, uranium--8.5 wt percent niobium, uranium--10 wt percent niobium, and uranium--7.5 wt percent niobium--2.5 wt percent zirconium), the hydrogen content of the tensile specimens, and the hydrogen gas pressure during the annealing at 850 0 C of the tensile test blanks prior to quenching were established. For these alloys, the tensile ductility decreases only slightly with increasing hydrogen content up to a critical hydrogen concentration above which the tensile ductility drops to nearly zero. The only alloy not displaying this sharp drop in tensile ductility was U--7.5 Nb--2.5 Zr, probably because sufficiently high hydrogen contents could not be achieved under our experimental arrangements. The critical hydrogen content for ductility loss increased with increasing hydrogen solubility in the alloy. Fracture surfaces produced by internal hydrogen embrittlement do not resemble those produced by stress corrosion cracking (SCC) in aqueous environments containing chloride ions. 8 figs

  14. Amorphous uranium alloy and use thereof

    International Nuclear Information System (INIS)

    Gambino, R.J.; McElfresh, M.W.; McGuire, T.R.; Plaskett, T.S.

    1991-01-01

    An amorphous alloy containing uranium and a member selected from the group N, P, As, Sb, Bi, S, Se, Te, Po and mixtures thereof; and use thereof for storage medium, light modulator or optical isolator. (author) figs

  15. Biaxial Testing of 2195 Aluminum Lithium Alloy Using Cruciform Specimens

    Science.gov (United States)

    Johnston, W. M.; Pollock, W. D.; Dawicke, D. S.; Wagner, John A. (Technical Monitor)

    2002-01-01

    A cruciform biaxial test specimen was used to test the effect of biaxial load on the yield of aluminum-lithium alloy 2195. Fifteen cruciform specimens were tested from 2 thicknesses of 2195-T8 plate, 0.45 in. and 1.75 in. These results were compared to the results from uniaxial tensile tests of the same alloy, and cruciform biaxial tests of aluminum alloy 2219-T87.

  16. Experimental study on uranium alloys for hydrogen storage

    International Nuclear Information System (INIS)

    Deaconu, M.; Meleg, T.; Dinu, A.; Mihalache, M.; Ciuca, I.; Abrudeanu, M.

    2013-01-01

    The heaviest isotope of hydrogen is one of critically important elements in the field of fusion reactor technology. Conventionally, uranium metal is used for the storage of heavier isotopes of hydrogen (D and T). Under appropriate conditions, uranium absorbs hydrogen to form a stable UH 3 compound when exposed to molecular hydrogen at the temperature range of 300-500 O C at varied operating pressure below one atmosphere. However, hydriding-dehydriding on pure uranium disintegrates the specimen into fine powder. The powder is highly pyrophoric and has low heat conductivity, which makes it difficult to control the temperature, and has a high possibility of contamination Due to the powdering effect as hydrogen in uranium, alloying uranium with other metal looks promising for the use of hydrogen storage materials. This paper has the aim to study the hydriding properties of uranium alloys, including U-Ti U-Mo and U-Ni. The uranium alloys specimens were prepared by melting the constituent elements by means of simultaneous measurements of thermo-gravimetric and differential thermal analyses (TGA-DTA) and studied in as cast condition as hydrogen storage materials. Then samples were thermally treated under constant flow of hydrogen, at various temperatures between 573-973 0 K. The structural and absorption properties of the products obtained were examined by thermo-gravimetric analysis (TG), X-ray diffraction (XRD) and scanning electron microscopy (SEM). They slowly reacted with hydrogen to form the ternary hydride and the hydrogenated samples mainly consisted of the pursued ternary hydride bat contained also U or UO 2 and some transient phase. (authors)

  17. Development of casting techniques for uranium and uranium alloys

    International Nuclear Information System (INIS)

    Singh, S.P.

    2003-01-01

    The casting process concerning furnace set-up, mould temperatures, pouring temperatures, out gassing, post heating, casting recovery and crucible and mould clean-up is discussed. Some applications of casting theory can be made in practice, but experience in handling the metal is most valuable in the successful solution of a new problem. The casting of uranium alloys using induction stirring of the melt to promote homogeneity in the casting is described. A few remarks are made concerning safety aspects associated with the casting of uranium

  18. Solidification microstructures of aluminium-uranium alloys

    International Nuclear Information System (INIS)

    Ambrozio Filho, F.; Vieira, R.R.

    1976-01-01

    The solidification of microstrutures of aluminium-uranium alloys in the range of 4 to 20% uranium is investigated. The solidification was obtained both in ingot molds and under controlled directional solidification. The conditions for the presence of primary crystals and eutectic are discussed and an analysis of the influence of variables (growth rate and thermal gradient in the liquid) on the alloy structure is made. The effect of cooling rate on the alloy structures has been determined. It is found that the resulting structure can be derived from the kinectics concept, as required by the coupled-zone theory. Suggestions on the qualitative intervals of composition and temperatures with eutectic growth are presented [pt

  19. Effect of passivation with CO on the electrochemical corrosion behavior of uranium-niobium alloy

    International Nuclear Information System (INIS)

    Fu Xiaoguo; Dai Lianxin; Zou Juesheng; Bai Chaomao; Wang Xiaolin

    2000-01-01

    Electrochemical studies are performed to investigate the corrosion resistance of uranium-niobium alloy before and after passivated with carbon monoxide. Using X-ray photoelectron spectroscopy (XPS), the surface composition of specimen passivated with carbon monoxide is determined. The corrosion resistance of uranium-niobium alloy is well improved because the passive layer (UC/UC x O y + Nb 2 O 5 + UO 2 ) on surface serves as passive film and increases the anodic impedance after the specimen is passivated with carbon monoxide

  20. Thermal stress relieving of dilute uranium alloys

    International Nuclear Information System (INIS)

    Eckelmeyer, K.H.

    1981-01-01

    The kinetics of thermal stress relieving of uranium - 2.3 wt % niobium, uranium - 2.0 wt % molybdenum, and uranium - 0.75 wt % titanium are reported and discussed. Two temperature regimes of stress relieving are observed. In the low temperature regime (T 0 C) the process appears to be controlled by an athermal microplasticity mechanism which can be completely suppressed by prior age hardening. In the high temperature regime (300 0 C 0 C) the process appears to be controlled by a classical diffusional creep mechanism which is strongly dependent on temperature and time. Stress relieving is accelerated in cases where it occurs simultaneously with age hardening. The potential danger of residual stress induced stress corrosion cracking of uranium alloys is discussed

  1. Welding of a powder metallurgy uranium alloy

    International Nuclear Information System (INIS)

    Holbert, R.K.; Doughty, M.W.; Alexander-Morrison, G.M.

    1989-01-01

    The interest at the Oak Ridge Y-12 Plant in powder metallurgy (P/M) uranium parts is due to the potential cost savings in the fabrication of the material, to achieving a more homogeneous product, and to the reduction of uranium scrap. The joining of P/M uranium-6 wt-% niobium (U-6Nb) alloys by the electron beam (EB) welding process results in weld porosity. Varying the EB welding parameters did not eliminate the porosity. Reducing the oxygen and nitrogen content in this P/M uranium material did minimize the weld porosity, but this step made the techniques of producing the material more difficult. Therefore, joining wrought and P/M U-6Nb rods with the inertia welding technique is considered. Since no gases will be evolved with the solid-state welding process and the weld area will be compacted, porosity should not be a problem in the inertia welding of uranium alloys. The welds that are evaluated are wrought-to-wrought, wrought-to-P/M, and P/M-to-P/M U-6Nb samples

  2. Thermal stress relieving of dilute uranium alloys

    International Nuclear Information System (INIS)

    Eckelmeyer, K.H.

    1980-01-01

    The kinetics of thermal stress relieving of uranium - 2.3 wt. % niobium, uranium - 2.0 wt. % molybdenum, and uranium - 0.75 wt. % titanium are reported and discussed. Two temperature regimes of stress relieving are observed. In the low temperature regime (T 0 C) the process appears to be controlled by an athermal microplasticity mechanism which can be completely suppressed by prior age hardening. In the high temperature regime (300 0 C 0 C) the process appears to be controlled by a classical diffusional creep mechanism which is strongly dependent on temperature and time. Stress relieving is accelerated in cases where it occurs simultaneously with age hardening. The potential danger of residual stress induced stress corrosion cracking of uranium alloys is discussed. It is shown that the residual stress relief which accompanies age hardening of uranium - 0.75% titanium more than compensates for the reduction in K/sub ISCC/ caused by aging. As a result, age hardening actually decreases the susceptibility of this alloy to residual stress induced stress corrosion cracking

  3. Progress Report on Alloy 617 Notched Specimen Testing

    Energy Technology Data Exchange (ETDEWEB)

    McMurtrey, Michael David [Idaho National Lab. (INL), Idaho Falls, ID (United States); Wright, Richard Neil [Idaho National Lab. (INL), Idaho Falls, ID (United States); Lillo, Thomas Martin [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-08-01

    Creep behavior of Alloy 617 has been extensively characterized to support the development of a draft Code Case to qualify Alloy 617 in Section III division 5 of the ASME Boiler and Pressure Vessel Code. This will allow use of Alloy 617 in construction of nuclear reactor components at elevated temperatures and longer periods of time (up to 950°C and 100,000 hours). Prior to actual use, additional concerns not considered in the ASME code need to be addressed. Code Cases are based largely on uniaxial testing of smooth gage specimens. In service conditions, components will generally be under multi axial loading. There is also the concern of the behavior at discontinuities, such as threaded components. To address the concerns of multi axial creep behavior and at geometric discontinuities, notched specimens have been designed to create conditions representative of the states that service components experience. Two general notch geometries have been used for these series of tests: U notch and V notch specimens. The notches produce a tri axial stress state, though not uniform across the specimen. Characterization of the creep behavior of the U notch specimens and the creep rupture behavior of the V notch specimens provides a good approximation of the behavior expected of actual components. Preliminary testing and analysis have been completed and are reported in this document. This includes results from V notch specimens tested at 900°C and 800°C. Failure occurred in the smooth gage section of the specimen rather than at the root of the notch, though some damage was present at the root of the notch, where initial stress was highest. This indicates notch strengthening behavior in this material at these temperatures.

  4. Thermodynamic properties of uranium in gallium–aluminium based alloys

    International Nuclear Information System (INIS)

    Volkovich, V.A.; Maltsev, D.S.; Yamshchikov, L.F.; Chukin, A.V.; Smolenski, V.V.; Novoselova, A.V.; Osipenko, A.G.

    2015-01-01

    Activity, activity coefficients and solubility of uranium was determined in gallium-aluminium alloys containing 1.6 (eutectic), 5 and 20 wt.% aluminium. Additionally, activity of uranium was determined in aluminium and Ga–Al alloys containing 0.014–20 wt.% Al. Experiments were performed up to 1073 K. Intermetallic compounds formed in the alloys were characterized by X-ray diffraction. Partial and excess thermodynamic functions of U in the studied alloys were calculated. - Highlights: • Thermodynamics of uranium is determined in Ga–Al alloys of various compositions. • Uranium in the mixed alloys interacts with both components, Ga and Al. • Interaction of U with Al increases with decreasing temperature. • Activity and solubility of uranium depend on Al content in Ga–Al alloys.

  5. Thermodynamic properties of uranium in gallium–aluminium based alloys

    Energy Technology Data Exchange (ETDEWEB)

    Volkovich, V.A., E-mail: v.a.volkovich@urfu.ru [Department of Rare Metals and Nanomaterials, Institute of Physics and Technology, Ural Federal University, Ekaterinburg, 620002 (Russian Federation); Maltsev, D.S.; Yamshchikov, L.F. [Department of Rare Metals and Nanomaterials, Institute of Physics and Technology, Ural Federal University, Ekaterinburg, 620002 (Russian Federation); Chukin, A.V. [Department of Theoretical Physics and Applied Mathematics, Institute of Physics and Technology, Ural Federal University, Ekaterinburg, 620002 (Russian Federation); Smolenski, V.V.; Novoselova, A.V. [Institute of High-Temperature Electrochemistry UD RAS, Ekaterinburg, 620137 (Russian Federation); Osipenko, A.G. [JSC “State Scientific Centre - Research Institute of Atomic Reactors”, Dimitrovgrad, 433510 (Russian Federation)

    2015-10-15

    Activity, activity coefficients and solubility of uranium was determined in gallium-aluminium alloys containing 1.6 (eutectic), 5 and 20 wt.% aluminium. Additionally, activity of uranium was determined in aluminium and Ga–Al alloys containing 0.014–20 wt.% Al. Experiments were performed up to 1073 K. Intermetallic compounds formed in the alloys were characterized by X-ray diffraction. Partial and excess thermodynamic functions of U in the studied alloys were calculated. - Highlights: • Thermodynamics of uranium is determined in Ga–Al alloys of various compositions. • Uranium in the mixed alloys interacts with both components, Ga and Al. • Interaction of U with Al increases with decreasing temperature. • Activity and solubility of uranium depend on Al content in Ga–Al alloys.

  6. Method of removing niobium from uranium-niobium alloy

    International Nuclear Information System (INIS)

    Pollock, E.N.; Schlier, D.S.; Shinopulos, G.

    1992-01-01

    This patent describes a method of removing niobium from a uranium-niobium alloy. It comprises dissolving the uranium-niobium alloy metal pieces in a first aqueous solution containing an acid selected from the group consisting of hydrochloric acid and sulfuric acid and fluoboric acid as a catalyst to provide a second aqueous solution, which includes uranium (U +4 ), acid radical ions, the acids insolubles including uranium oxides and niobium oxides; adding nitric acid to the insolubles to oxidize the niobium oxides to yield niobic acid and to complete the solubilization of any residual uranium; and separating the niobic acid from the nitric acid and solubilized uranium

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

    International Nuclear Information System (INIS)

    Hemperly, V.C.

    1975-01-01

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

  8. Hot rolling of thick uranium molybdenum alloys

    Science.gov (United States)

    DeMint, Amy L.; Gooch, Jack G.

    2015-11-17

    Disclosed herein are processes for hot rolling billets of uranium that have been alloyed with about ten weight percent molybdenum to produce cold-rollable sheets that are about one hundred mils thick. In certain embodiments, the billets have a thickness of about 7/8 inch or greater. Disclosed processes typically involve a rolling schedule that includes a light rolling pass and at least one medium rolling pass. Processes may also include reheating the rolling stock and using one or more heavy rolling passes, and may include an annealing step.

  9. Spectrographic analysis of uranium-molybdenum alloys

    International Nuclear Information System (INIS)

    Roca, M.

    1967-01-01

    A spectrographic method of analysis has been developed for uranium-molybdenum alloys containing up to 10 % Mo. The carrier distillation technique, with gallium oxide and graphite as carriers, is used for the semiquantitative determination of Al, Cr, Fe, Ni and Si, involving the conversion of the samples into oxides. As a consequence of the study of the influence of the molybdenum on the line intensities, it is useful to prepare only one set of standards with 0,6 % MoO 3 . Total burning excitation is used for calcium, employing two sets of standards with 0,6 and 7.5 MoO 3 . (Author) 5 refs

  10. 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)

  11. Atmospheric corrosion of uranium-carbon alloys

    International Nuclear Information System (INIS)

    Rousset, P.; Accary, A.

    1965-01-01

    The authors study the corrosion of uranium-carbon alloys having compositions close to that of the mono-carbide; they show that the extent of the observed corrosion effects increases with the water vapour content of the surrounding gas and they conclude that the atmospheric corrosion of these alloys is due essentially to the humidity of the air, the effect of the oxygen being very slight at room temperature. They show that the optimum conditions for preserving U-C alloys are either a vacuum or a perfectly dry argon atmosphere. The authors have also established that the type of corrosion involved is a corrosion which 'cracks under stress' and is transgranular (it can also be intergranular in the case of sub-stoichiometric alloys). They propose, finally, two hypotheses for explaining this mechanism, one of which is illustrated by the existence, at the fissure interface, of corrosion products which can play the role of 'corners' in the mono-carbide grains. (authors) [fr

  12. X-ray topography of uranium alloys

    International Nuclear Information System (INIS)

    Le Naour, L.

    1984-01-01

    The limitations of x-ray topography methods are due to the variety of structures studied and to the variation of the amplitude of the scattering of incident beams. It is difficult to evaluate the aberrations and the imperfections of the material studied. Interpretation of the x-ray images will often be delicate and that is aggravated by the complexity of the diffraction spectrum of uranium. This negative aspect is compensated for by the advantage that chemical or electrochemical preparations of the alloy surface, along with alterations that can take place and the lack of trueness are avoided. Precise and very reproducible numerical data can be derived from the patterns. The structure of alloys, at a given scale, is revealed and characterized by quantitative parameters such as size of grains or sub-grains, dispersion of their dimensions, mutual disorientations and the continuous or discontinuous nature of the latter. The results of this research, therefore, justify the use of methods inspired by the Berg-Barrett technique. These diffraction procedures constitute a useful means for investigating many elements of microstructure that closely govern the behavior under irradiation of the materials being examined

  13. Characterization of the uranium--2 weight percent molybdenum alloy

    International Nuclear Information System (INIS)

    Hemperly, V.C.

    1976-01-01

    The uranium-2 wt percent molybdenum alloy was prepared, processed, and age hardened to meet a minimum 930-MPa yield strength (0.2 percent) with a minimum of 10 percent elongation. These mechanical properties were obtained with a carbon level up to 300 ppM in the alloy. The tensile-test ductility is lowered by the humidity of the laboratory atmosphere

  14. Thermal Cycling of Uranium Dioxide - Tungsten Cermet Fuel Specimens

    Energy Technology Data Exchange (ETDEWEB)

    Gripshover, P.J.; Peterson, J.H.

    1969-12-08

    In phase I tungsten clad cermet fuel specimens were thermal cycled, to study the effects of fuel loading, fuel particle size, stablized fuel, duplex coatings, and fabrication techniques on dimensional stability during thermal cycling. In phase II the best combination of the factors studies in phase I were combined in one specimen for evaluation.

  15. Equations of state for enriched uranium and uranium alloy to 3500 MPa

    International Nuclear Information System (INIS)

    Bai Chaomao; Hai Yuying; Liu Jenlong; Li Zhenrong

    1990-04-01

    The volume compressions of 6 kinds of cast materials including enriched uranium, poor uranium, U-0.57 wt% Ti, U-0.33 wt% Nb, U-2.85 wt% Nb and U-7.5 wt% Nb-3.3 wt% Zr have been determined by monitoring piston displacements in a piston cylinder apparatus with double strengthening rings to 3500 MPa at room temperature. The dilation of the cylinder vessel and the press deformation were corrected by some experiments. The calculational data free from using the standard sample closed with used standard sample. The volume compressions of enriched uranium and poor uranium are nearly coincident. Pure uranium is more compressible than uranium alloys. These values of enriched uranium are in close agreement with values of Bridgman's pure uranium. The fitting coefficients of Bridgman's polynomial and Anderson's equation of state and isothermal bulk modules for the above materials are given

  16. Track Detection Technique Using CR-39 for Determining Depleted Uranium in Biological Specimens

    International Nuclear Information System (INIS)

    Murbat, S.M.

    2013-01-01

    Track detecting technique using CR-39 track detector has been implemented for determining depleted uranium concentration in biological specimens (tissues, bones, and blood) of patients infected with cancer diseases. Results were compared with specimens of patients infected with conventional diseases (noncancerous). Specimens were collected from middle and south of Iraq have been contaminated with depleted uranium in the Gulf war in 1991. Results show that this technique is efficient for determining depleted uranium concentration in biological specimens. It was found that all studies samples determine for patients infected with cancer diseases contain a high concentration of depleted uranium (more than the international standard) comparing with noncancerous diseases. Moreover, it was found that persons infected with Leukemia show more sensitive to uranium concentrations to induce the diseases (66-202 ppb), while (116- 1910 ppb) concentrations were needed for inducing cancer diseases in organs and tissues. Result confirmed the correlation between cancerous diseases and the munitions made of depleted uranium used in the Gulf war in 1991 leads to contaminate the Iraqi environment and causes a high risk against people in Iraq.

  17. Shape memory effects in a uranium + 14 at. % niobium alloy

    International Nuclear Information System (INIS)

    Vandermeer, R.A.; Ogle, J.C.; Snyder, W.B. Jr.

    1978-01-01

    There is a class of alloys that, on cooling from elevated temperatures, experience a martensitic phase change. Some of these, when stressed in the martensitic state to an apparently plastic strain, recover their predeformed shape simply by heating. This striking shape recovery is known as the ''shape memory effect'' (SME). Up to a certain limiting strain, epsilon/sub L/, 100% shape recovery may be accomplished. This memory phenomenon seems to be attributable to the thermoelastic nature of and deformational modes associated with the phase transformation in the alloy. Thus, shape recovery results when a stress-biased martensite undergoes a heat-activated reversion back to the parent phase from which it originated. There are uranium alloys that demonstrate SME-behavior. Uranium-rich, uranium--niobium alloys were the first to be documented; New experimental observations of SME in a polycrystalline uranium--niobium alloy are presented. This alloy can exhibit a two-way memory under cetain circumstances. Additional indirect evidence is presented suggesting that the characteristics of the accompanying phase transformation in this alloy meet the criteria or ''selection rules'' deemed essential for SME

  18. Study of the pyrophoric characteristics of uranium-iron alloys

    International Nuclear Information System (INIS)

    Duplessis, X.

    2000-01-01

    The objective of the study is to understand the pyrophoric characteristics of uranium-iron alloys. In order to carry out this research we have elected to use uranium-iron alloy powder with granules of 200 μm and 1000 μm diameter with 4%, 10.8% and 14% iron content. The experiments were performed on small samples of few milligrams and on larger quantities of few hundred grams. The main conclusions obtained are the followings: -The reaction start at 453 K (180 deg. C) and the ignition at 543 K (270 deg. C) - The influence of the specific area seems more important than the iron concentration in the alloys - When the alloy ignites, the fire spreads quickly and the alloy rapidly consumes. (author)

  19. Alloys of uranium and aluminium with low aluminium content

    International Nuclear Information System (INIS)

    Cabane, G.; Englander, M.; Lehmann, J.

    1955-01-01

    Uranium, as obtained after spinning in phase γ, presents an heterogeneous structure with large size grains. The anisotropic structure of the metal leads to an important buckling and surface distortion of the fuel slug which is incompatible with its tubular cladding for nuclear fuel uses. Different treatments have been made to obtain an isotropic structure presenting high thermal stability (laminating, hammering and spinning in phase α) without success. Alloys of uranium and aluminium with low aluminium content present important advantage in respect of non allied uranium. The introduction of aluminium in the form of intermetallic compound (UAl 2 ) gives a better resistance to thermal fatigue. Alloys obtained from raw casting present an improved buckling and surface distortion in respect of pure uranium. This improvement is obtained with uranium containing between 0,15 and 0,5 % of aluminium. An even more improvement in thermal stability is obtained by thermal treatments of these alloys. These new characteristics are explained by the fine dispersion of the UAl 2 particles in uranium. The results after treatments obtained from an alloy slug containing 0,4 % of aluminium show no buckling or surface distortion and no elongation. (M.P.)

  20. Fabrication of uranium carbide/beryllium carbide/graphite experimental-fuel-element specimens

    International Nuclear Information System (INIS)

    Muenzer, W.A.

    1978-01-01

    A method has been developed for fabricating uranium carbide/beryllium carbide/graphite fuel-element specimens for reactor-core-meltdown studies. The method involves milling and blending the raw materials and densifying the resulting blend by conventional graphite-die hot-pressing techniques. It can be used to fabricate specimens with good physical integrity and material dispersion, with densities of greater than 90% of the theoretical density, and with a uranium carbide particle size of less than 10 μm

  1. Effect of laser power and specimen temperature on atom probe analyses of magnesium alloys

    International Nuclear Information System (INIS)

    Oh-ishi, K.; Mendis, C.L.; Ohkubo, T.; Hono, K.

    2011-01-01

    The influence of laser power, wave length, and specimen temperature on laser assisted atom probe analyses for Mg alloys was investigated. Higher laser power and lower specimen temperature led to improved mass and spatial resolutions. Background noise and mass resolutions were degraded with lower laser power and higher specimen temperature. By adjusting the conditions for laser assisted atom probe analyses, atom probe results with atomic layer resolutions were obtained from all the Mg alloys so far investigated. Laser assisted atom probe investigations revealed detailed chemical information on Guinier-Preston zones in Mg alloys. -- Research highlights: → We study performance of UV laser assisted atom probe analysis for Mg alloys. → There is an optimized range of laser power and specimen temperature. → Optimized UV laser enables atom probe data of Mg alloys with high special resolution.

  2. Study on segregation of aluminium-uranium alloys

    International Nuclear Information System (INIS)

    Lima, Rui Marques de

    1979-01-01

    The relations between alloy solidification and solute segregation were considered. The solidification structure and the solute redistribution during the solidification of alloys with dendritic micro morphology were studied. The macro and micro segregation theories were reviewed. The mechanisms that could change the solidification structure were taken into account in the context of more homogeneous alloy production. Aluminum alloys solidification structures and segregation were studied experimentally in the 13 to 45% uranium range, usually considering solidification in static molds. The uranium alloys with up to 20% uranium were studied both for solidification in ingot molds and for controlled directional solidification. It was verified that these alloy compositions had structures similar to those of hipoeutectic alloys, showing an a phase with dendritic morphology and inter dendritic eutectic. For the alloys with more than 25% uranium, it was observed the formation of UAl 3 and UAl 4 phases with dendritic morphology. The dendritic UAl 3 , phase morphology was affected both by the solute concentration in the alloy and by the growth rate. The dendritic UAl 3 phase non-singular aspect could be destroyed with decrease of the alloy solute concentration. In the alloys obtained with higher cooling rates it was found a tendency for the formation of substantial quantities of equi axial crystals of the solute enriched phases in the central regions of the ingot upper half. In the more external regions it was observed dendritic growth of these phases, for alloy compositions with over 25% uranium. An adequate reduction in the cooling rate changed the solidification structure form and distribution, as well as the segregation type and intensity. The uranium content in the solidified macro structures is presented as a function of: cooling rate, superheating, mold size, mold form and its temperature, number of remelting and time for the melt homogenization and agitation. It was

  3. Metallurgical processing of the uranium-0.75 titanium alloy

    International Nuclear Information System (INIS)

    Jessen, N.C.

    1976-01-01

    Although the addition of titanium is an effective means of strengthening uranium, careful control of casting, homogenization, and heat treatment are necessary to optimize mechanical properties. Quenching of the alloy provides increased strength and elongation; however, subsequent low temperature aging will increase the strength even higher at the sacrifice of ductility. The properties of the alloy are quench rate sensitive and quenching produces high residual stresses in the alloy. The residual stresses can be reduced by mechanical deformation with only slight degradation of the mechanical properties. 15 figures

  4. Some potential strategies for the treatment of waste uranium metal and uranium alloys

    International Nuclear Information System (INIS)

    Burns, C.J.; Frankcom, T.M.; Gordon, P.L.; Sauer, N.N.

    1993-01-01

    Large quantities of uranium metal chips and turnings stored throughout the DOE Complex represent a potential hazard, due to the reactivity of this material toward air and water. Methods are being sought to mitigate this by conversion of the metal, via room temperature solutions routes, to a more inert oxide form. In addition, the recycling of uranium and concomitant recovery of alloying metals is a desirable goal. The emphasis of the authors' research is to explore a variety of oxidation and reduction pathways for uranium and its compounds, and to investigate how these reactions might be applied to the treatment of bulk wastes

  5. Effect of aging on the general corrosion and stress corrosion cracking of uranium--6 wt % niobium alloy

    International Nuclear Information System (INIS)

    Koger, J.W.; Ammons, A.M.; Ferguson, J.E.

    1975-11-01

    Mechanical properties of the uranium-6 wt percent niobium alloy change with aging time and temperature. In general, the ultimate tensile strength and hardness reach a peak, while elongation becomes a minimum at aging temperatures between 400 and 500 0 C. The first optical evidence of a second phase was in the 400 0 C-aged alloy, while complete transformation to a two-phase structure was seen in the 600 0 C-aged alloy. The maximum-strength conditions correlate with the minimum stress corrosion cracking (SCC) resistance. The maximum SCC resistance is found in the as-quenched and 150, 200, and 600 0 C-aged specimens. The as-quenched and 300 0 C-aged specimens had the greatest resistance to general corrosion in aqueous chloride solutions; the 600 0 C-aged specimen had the least resistance

  6. Durability of adhesive bonds to uranium alloys, tungsten, tantalum, and thorium

    International Nuclear Information System (INIS)

    Childress, F.G.

    1975-01-01

    Long-term durability of epoxy bonds to alloys of uranium (U-Nb and Mulberry), nickel-plated uranium, thorium, tungsten, tantalum, tantalum--10 percent tungsten, and aluminum was evaluated. Significant strengths remain after ten years of aging; however, there is some evidence of bond deterioration with uranium alloys and thorium stored in ambient laboratory air

  7. Dissolution of metallic uranium and its alloys. Part 1. Review of analytical and process-scale metallic uranium dissolution

    International Nuclear Information System (INIS)

    Laue, C.A.; Gates-Anderson, D.; Fitch, T.E.

    2004-01-01

    This review focuses on dissolution/reaction systems capable of treating uranium metal waste to remove its pyrophoric properties. The primary emphasis is the review of literature describing analytical and production-scale dissolution methods applied to either uranium metal or uranium alloys. A brief summary of uranium's corrosion behavior is included since the corrosion resistance of metals and alloys affects their dissolution behavior. Based on this review, dissolution systems were recommended for subsequent screening studies designed to identify the best system to treat depleted uranium metal wastes at Lawrence Livermore National Laboratory (LLNL). (author)

  8. Corrosion and protection of uranium alloy penetrators

    International Nuclear Information System (INIS)

    Weirick, L.J.; Johnson, H.R.; Dini, J.W.

    1975-06-01

    Penetrators made from either a U--3/4 percent Ti alloy or a U--3/4 percent Mo--3/4 percent Zr--3/4 percent Nb--1/2 percent Ti alloy (''Quad'') corrode mildly in moist air, significantly in moist nitrogen, and severely in salt fog. Adequate protection was provided in moist air and nitrogen by coating with electroplated nickel, electroplated nickel and zinc with a chromate finish, and galvanized zinc with a chromate finish. In salt fog, electroplated nickel offered only temporary protection whereas galvanized zinc and electroplated nickel-zinc provided long-lasting protection. The resistance of uncoated penetrators was affected variously by dissimilar metal couplings. Aluminum protected the Quad alloy and adversely affected the U--3/4 percent Ti alloy, whereas steel enhanced localized corrosion in both. (U.S.)

  9. Low content uranium alloys for nuclear fuels

    International Nuclear Information System (INIS)

    Aubert, H.; Laniesse, J.

    1964-01-01

    A description is given of the structure and the properties of low content alloys containing from 0.1 to 0.5 per cent by weight of Al, Fe, Cr, Si, Mo or a combination of these elements. A study of the kinetics and of the mode of transformation has made it possible to choose the most satisfactory thermal treatment. An attempt has been made to prepare alloys suitable for an economical industrial development having a small α grain structure without marked preferential orientation, with very fine and stable precipitates as well as a high creep-resistance. The physical properties and the mechanical strength of these alloys are given for temperatures of 20 to 600 deg C. These alloys proved very satisfactory when irradiated in the form of normal size fuel elements. (authors) [fr

  10. Determination of uranium in fissium-uranium alloy and in fissium dross

    International Nuclear Information System (INIS)

    Bodnar, L.Z.

    1976-01-01

    Dissolution and analysis techniques for fissium-uranium alloy and fissium dross are described. The fuming technique of dissolution effectively eliminated all interferring elements in the titration determination of U. The results from the semiquantitative analysis of fission dross by spark source mass spectrometry were tabulated

  11. Uranium alloys for using in fast breeder reactors

    International Nuclear Information System (INIS)

    Moura Neto, C.; Pires, O.S.

    1988-08-01

    The U-Zr and U-Ti alloys are studied, given emphasis to the high solute solubility in gamma phase of uranium, which is suitable for using as metal fuel in fast breeder reactors. The alloys were prepared in electron beam furnaces and submitted to X-ray diffraction, X-ray fluorescence, microhardness, optical metallography, and chemical analysis. The obtained values are good agreements with the literature data. The study shows that the U-Zr presents better characteristics than the U-Ti for using as fuel in fast breeder reactors. (M.C.K.) [pt

  12. Metallurgical structures in a high uranium-silicon alloy

    International Nuclear Information System (INIS)

    Wyatt, B.S.; Berthiaume, L.C.; Conversi, J.L.

    1968-10-01

    The effects of fabrication and heat treatment variables on the structure of a uranium -- 3.96 wt% silicon alloy have been studied using optical microscopy, quantitative metallography and hardness determinations. It has been shown that an optimum temperature exists below the peritectoid temperature where the maximum amount of transformation to U 3 Si occurs in a given period of time. The time required to fully transform an as-cast alloy at this optimum temperature is affected by the size of the primary U 3 Si 2 dendrites. With a U 3 Si 2 particle size of <12 μm complete transformation can be achieved in four hours. (author)

  13. Contribution to the micrographic study of uranium and its alloys

    International Nuclear Information System (INIS)

    Monti, H.

    1956-06-01

    The present report is the result of research carried out by the radio metallurgy section, to perfect micrographic techniques applicable to the study of samples of irradiated uranium. In the first part of this work, two polishing baths are developed, having the qualities with a minimum of disadvantages inherent in their respective compositions: they are, on the one hand perchloric acid-ethanol mixtures, and on the other hand a phospho-chromic-ethanol bath. In the chapter following, the micrographic attack of uranium is studied. The only satisfactory process is oxidation by cathode bombardment forming epitaxic layers. In the third chapter, an attempt is made to characterise the different surface states of the uranium by dissolution potential measurements and electronic diffraction. In the fourth chapter are given some examples of the application of these techniques to the micrographic study of various uranium alloys. In an appendix, it is shown how the chemical oxidation after phospho-chromic-alcohol polishing allows the different inclusions present in the molten uranium to be distinguished. By X-ray diffraction, uranium monocarbide and mononitride inclusions in particular are characterised. (author) [fr

  14. X-ray diffraction (XRD) characterization of microstrain in some iron and uranium alloys

    International Nuclear Information System (INIS)

    Kimmel, G.; Dayan, D.; Frank, G.A.; Landau, A.

    1996-01-01

    The high linear attenuation coefficient of steel, uranium and uranium based alloys is associated with the small penetration depth of X-rays with the usual wavelength used for diffraction. Nevertheless, by using the proper surface preparation technique, it is possible of obtaining surfaces with bulk properties (free of residual mechanical microstrain). Taking advantage of the feasibility to obtain well prepared surfaces, extensive work has been conducted in studying XRD line broadening effects from flat polycrystalline samples of steel, uranium and uranium alloys

  15. Nuclear criticality safety parameter evaluation for uranium metallic alloy

    Energy Technology Data Exchange (ETDEWEB)

    Sanchez, Andrea; Abe, Alfredo, E-mail: andreasdpz@hotmail.com, E-mail: abye@uol.com.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil). Centro de Energia Nuclear

    2013-07-01

    Nuclear criticality safety during fuel fabrication process, transport and storage of fissile and fissionable materials requires criticality safety analysis. Normally the analysis involves computer calculations and safety parameters determination. There are many different Criticality Safety Handbooks where such safety parameters for several different fissile mixtures are presented. The handbooks have been published to provide data and safety principles for the design, safety evaluation and licensing of operations, transport and storage of fissile and fissionable materials. The data often comprise not only critical values, but also subcritical limits and safe parameters obtained for specific conditions using criticality safety calculation codes such as SCALE system. Although many data are available for different fissile and fissionable materials, compounds, mixtures, different enrichment level, there are a lack of information regarding a uranium metal alloy, specifically UMo and UNbZr. Nowadays uranium metal alloy as fuel have been investigated under RERTR program as possible candidate to became a new fuel for research reactor due to high density. This work aim to evaluate a set of criticality safety parameters for uranium metal alloy using SCALE system and MCNP Monte Carlo code. (author)

  16. Prediction of multiaxial fatigue life for notched specimens of titanium alloy TC4

    Energy Technology Data Exchange (ETDEWEB)

    Wu, Z. R.; Li, Z. X. [Southeast University, Nanjing (China); Hu, X. T.; Song, Y. D. [Nanjing University, Nanjing (China)

    2016-05-15

    Both the proportional and nonproportional multiaxial fatigue tests were conducted on two kinds of notched specimens of titanium alloy TC4. The multiaxial fatigue critical area of notched specimen is considered as the location experiencing the maximum damage. It is unsatisfactory to predict the multiaxial fatigue life with the local stress and strain in the fatigue critical area. The critical distance concepts are employed in the multiaxial life prediction method for notched specimens. The proposed method was checked by the test data of TC4 notched specimens. The prediction results are almost within a factor of three scatter band of the test results.

  17. A method for the electrolytic coating of uranium or uranium alloy parts, and parts thus obtained

    International Nuclear Information System (INIS)

    1973-01-01

    A method, preceded by a surface treatment, for applying an electrolytic coating (e.g. of nickel) on uranium, or uranium alloy parts. This method is characterized in that the previous surface treatment comprises a chemical removal of grease in halogenated solvent bath (free from halogen ions) and an anodic scouring in a buffered aqueous solution solution of an acid free from halogen ions. The coating can be applied to fuel elements for nuclear industry, counter-weight for aeronautics and space industries and to radiation shields [fr

  18. Examination of temperature-induced shape memory of uranium--5.3-to 6.9 weight percent niobium alloys

    International Nuclear Information System (INIS)

    Hemperly, V.C.

    1976-01-01

    The uranium-niobium alloy system was examined in the range of 5.3-to-6.9 weight percent niobium with respect to shape memory, mechanical properties, metallography, Coefficients of linear thermal expansion, and differential thermal analysis. Shape memory increased with increasing niobium levels in the study range. There were no useful correlations found between shape memory and the other tests. Coefficients of linear thermal expansion tests of as-quenched 5.8 and 6.2 weight percent niobium specimens, but not 5.3 and 6.9 weight percent niobium specimens, had a contraction component on heating, but the phenomenon was not a contributor to shape memory

  19. Detection of thallium and uranium in well water and biological specimens of an eastern Croatian population.

    Science.gov (United States)

    Curković, Mario; Sipos, Laszlo; Puntarić, Dinko; Dodig-Ćurković, Katarina; Pivac, Nela; Kralik, Kristina

    2013-09-01

    Abstract Using inductively-coupled plasma mass spectrometry (ICP-MS), we measured the concentrations of thallium and uranium in local water resources from three villages (Ćelije, Draž, and Potnjani) in eastern Croatia, with the aim to determine if they were associated with the levels of these same elements in the serum, urine, and hair collected from the residents of this area. The exposure of the local population to thallium and uranium through drinking water was generally low. ICP-MS was capable of measuring the levels of both of the elements in almost all of the analysed samples. Although there were differences in the concentrations of both elements in water samples and biological specimens taken from the residents, they did not reach the maximum contaminant level in any of the four sample types studied. Although hair was previously reported as an excellent indicator of occupational and environmental exposure to various elements, our study did not confirm it as a reliable biological material for tracing thallium and uranium levels, mainly due to the very low concentrations of these elements, often well below the detection limit. However, our results have shown that the concentration of thallium and uranium in drinking water can be effectively traced in urine samples.

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

    International Nuclear Information System (INIS)

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

    1987-01-01

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

  1. Some properties of aluminum-uranium alloys in the cast, rolled and annealed conditions

    International Nuclear Information System (INIS)

    Jones, T.I.; McGee, I.J.; Norlock, L.R.

    1960-06-01

    The metallographic and hardness changes associated with the rolling and subsequent. annealing of aluminum alloys containing up to 30-wt.% uranium have been described. The alloys possessed good rolling properties. However the richer alloys were unusual in that after an initial reduction,, further cold rolling caused softening. In the alloy range examined, increasing uranium contents caused reduced preferred orientation. Qualitative explanations have been proposed to account for the observations on roll softening and preferred orientation. Heat-treating and ageing experiments confirmed that the solid solubility of uranium in aluminum is negligible. (author)

  2. Metallurgical examination of powder metallurgy uranium alloy welds

    International Nuclear Information System (INIS)

    Morrison, A.G.M.; Dobbins, A.G.; Holbert, R.K.; Doughty, M.W.

    1986-01-01

    Inertia welding provided a successful technique for joining full density, powder metallurgy uranium-6 wt pct niobium alloy. Initial joining attempts concentrated on the electron beam method, but this method failed to produce a sound weld. The electron beam welds and the inertia welds were evaluated by radiography and metallography. Electron beam welds were attempted on powder metallurgy plates which contained various levels of oxygen and nitrogen. All welds were porous. Sixteen inertia welds were made and all welds were radiographically sound. The tensile properties of the joints were found to be equivalent to the p/m base metal properties

  3. Spectrographic determination of niobium in uranium - niobium alloys

    International Nuclear Information System (INIS)

    Charbel, M.Y.; Lordello, A.R.

    1984-01-01

    A method for the spectrographic determination of niobium in uranium-niobium alloys in the concentration range 1-10% has been developed. The metallic sample is converted to oxide by calcination in a muffle furnace at 800 0 C for two hours. The standards are prepared synthetically by dry-mixing. One part of the sample or standard is added to nineteen parts of graphite powder and the mixture is excited in a DC arc. Hafnium has been used as internal standard. The precision of the method is + - 4.8%. (Author) [pt

  4. Corrosive and cytotoxic properties of compact specimens and microparticles of Ni-Cr dental alloy.

    Science.gov (United States)

    Ristic, Ljubisa; Vucevic, Dragana; Radovic, Ljubica; Djordjevic, Snezana; Nikacevic, Milutin; Colic, Miodrag

    2014-04-01

    Nickel-chromium (Ni-Cr) dental alloys have been widely used in prosthodontic practice, but there is a permanent concern about their biocompatibility due to the release of metal ions. This is especially important when Ni-Cr metal microparticles are incorporated into gingival tissue during prosthodontic procedures. Therefore, the aim of this study was to examine and compare the corrosion and cytotoxic properties of compact specimens and microparticles of Ni-Cr dental alloy. Ni-Cr alloy, Remanium CSe bars (4 mm diameter), were made by the standard casting method and then cut into 0.5-mm-thick disks. Metal particles were obtained by scraping the bars using a diamond instrument for crown preparation. The microstructure was observed by an optical microscope. Quantitative determination and morphological and dimensional characterization of metal particles were carried out by a scanning electron microscope and Leica Application Suite software for image analysis. Corrosion was studied by conditioning the alloy specimens in the RPMI 1640 medium, containing 10% fetal calf serum in an incubator with 5% CO2 for 72 hours at 37°C. Inductively coupled plasma-optical emission spectrometry was used to assess metal ion release. The cytotoxity of conditioning medium (CM) was investigated on L929 cells using an MTT test. One-way ANOVA was used for statistical analysis. After casting, the microstructure of the Remanium CSe compact specimen composed of Ni, Cr, Mo, Si, Fe, Al, and Co had a typical dendritic structure. Alloy microparticles had an irregular shape with a wide size range: from less than 1 μm to more than 100 μm. The release of metal ions, especially Ni and Mo from microparticles, was significantly higher, compared to the compact alloy specimen. The CM prepared from compact alloy was not cytotoxic at any tested dilutions, whereas CM from alloy microparticles showed dose-dependent cytotoxicity (90% CM and 45% CM versus control; p alloy. This could affect health on long

  5. Spectrographic analysis of uranium-based alloys; Analyse spectrographique d'alliages a base d'uranium

    Energy Technology Data Exchange (ETDEWEB)

    Baudin, G.; Blum, P.

    1959-07-01

    The authors describe a spectrographic method for dosing cobalt in cobalt-uranium alloys with cobalt content from 0.05 to 10 per cent. They describe sample preparation, alloy solution, spectrographic conditions, and photometry operations. In a second part, they address the dosing of boron in uranium borides. They implement the so-called 'porous cup' method. Boride is dissolved by fusion with Co{sub 3}-NaK [French] Uranium-Cobalt: il est decrit une methode spectrographique de dosage de cobalt dans des alliages cobalt-uranium pour des teneurs de 0,05 pour cent a 10 pour cent de Co. On opere sur solution avec le fer comme standard interne. Borure d'Uranium: ici encore on opere par la methode dite 'porous cup', le fer etant conserve comme standard interne. Le borure est mis en solution par fusion avec Co{sub 3}NaK. (auteurs)

  6. Kr ion irradiation study of the depleted-uranium alloys

    Science.gov (United States)

    Gan, J.; Keiser, D. D.; Miller, B. D.; Kirk, M. A.; Rest, J.; Allen, T. R.; Wachs, D. M.

    2010-12-01

    Fuel development for the reduced enrichment research and test reactor (RERTR) program is tasked with the development of new low enrichment uranium nuclear fuels that can be employed to replace existing high enrichment uranium fuels currently used in some research reactors throughout the world. For dispersion type fuels, radiation stability of the fuel-cladding interaction product has a strong impact on fuel performance. Three depleted-uranium alloys are cast for the radiation stability studies of the fuel-cladding interaction product using Kr ion irradiation to investigate radiation damage from fission products. SEM analysis indicates the presence of the phases of interest: U(Al, Si) 3, (U, Mo)(Al, Si) 3, UMo 2Al 20, U 6Mo 4Al 43 and UAl 4. Irradiations of TEM disc samples were conducted with 500 keV Kr ions at 200 °C to ion doses up to 2.5 × 10 19 ions/m 2 (˜10 dpa) with an Kr ion flux of 10 16 ions/m 2/s (˜4.0 × 10 -3 dpa/s). Microstructural evolution of the phases relevant to fuel-cladding interaction products was investigated using transmission electron microscopy.

  7. Kr ion irradiation study of the depleted-uranium alloys

    Energy Technology Data Exchange (ETDEWEB)

    Gan, J., E-mail: Jian.Gan@inl.go [Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-6188 (United States); Keiser, D.D. [Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-6188 (United States); Miller, B.D. [University of Wisconsin, 1500 Engineering Drive, Madison, WI 53706 (United States); Kirk, M.A.; Rest, J. [Argonne National Laboratory, 9700 South Cass Ave., Argonne, IL 60439 (United States); Allen, T.R. [University of Wisconsin, 1500 Engineering Drive, Madison, WI 53706 (United States); Wachs, D.M. [Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-6188 (United States)

    2010-12-01

    Fuel development for the reduced enrichment research and test reactor (RERTR) program is tasked with the development of new low enrichment uranium nuclear fuels that can be employed to replace existing high enrichment uranium fuels currently used in some research reactors throughout the world. For dispersion type fuels, radiation stability of the fuel-cladding interaction product has a strong impact on fuel performance. Three depleted-uranium alloys are cast for the radiation stability studies of the fuel-cladding interaction product using Kr ion irradiation to investigate radiation damage from fission products. SEM analysis indicates the presence of the phases of interest: U(Al, Si){sub 3}, (U, Mo)(Al, Si){sub 3}, UMo{sub 2}Al{sub 20}, U{sub 6}Mo{sub 4}Al{sub 43} and UAl{sub 4}. Irradiations of TEM disc samples were conducted with 500 keV Kr ions at 200 {sup o}C to ion doses up to 2.5 x 10{sup 19} ions/m{sup 2} ({approx}10 dpa) with an Kr ion flux of 10{sup 16} ions/m{sup 2}/s ({approx}4.0 x 10{sup -3} dpa/s). Microstructural evolution of the phases relevant to fuel-cladding interaction products was investigated using transmission electron microscopy.

  8. Development, preparation and characterization of uranium molybdenum alloys for dispersion fuel application

    International Nuclear Information System (INIS)

    Sinha, V.P.; Prasad, G.J.; Hegde, P.V.; Keswani, R.; Basak, C.B.; Pal, S.; Mishra, G.P.

    2009-01-01

    Most of the research and test reactors worldwide have undergone core conversion from high enriched uranium base fuel to low enriched uranium base fuel under the Reduced Enrichment for Research and Test Reactor (RERTR) program, which was launched in the late 1970s to reduce the risk of nuclear proliferation. To realize this goal, high density uranium compounds and γ-stabilized uranium alloy powder were identified. In Metallic Fuels Division of BARC, R and D efforts are on to develop these high density uranium base alloys. This paper describes the preparation flow sheet for different compositions of Uranium and molybdenum alloys by an innovative powder processing route with uranium and molybdenum metal powders as starting materials. The same composition of U-Mo alloys were also fabricated by conventional method i.e. ingot metallurgy route. The U-Mo alloys prepared by both the methods were then characterized by XRD for phase analysis. The photomicrographs of alloys with different compositions prepared by powder metallurgy and ingot metallurgy routes are also included in the paper. The paper also covers the comparison of properties of the alloys prepared by powder metallurgy and ingot metallurgy routes

  9. Development, preparation and characterization of uranium molybdenum alloys for dispersion fuel application

    Energy Technology Data Exchange (ETDEWEB)

    Sinha, V.P. [Metallic Fuels Division, Bhabha Atomic Research Centre, Mumbai 400085 (India)], E-mail: vedsinha@barc.gov.in; Prasad, G.J.; Hegde, P.V.; Keswani, R.; Basak, C.B.; Pal, S.; Mishra, G.P. [Metallic Fuels Division, Bhabha Atomic Research Centre, Mumbai 400085 (India)

    2009-04-03

    Most of the research and test reactors worldwide have undergone core conversion from high enriched uranium base fuel to low enriched uranium base fuel under the Reduced Enrichment for Research and Test Reactor (RERTR) program, which was launched in the late 1970s to reduce the risk of nuclear proliferation. To realize this goal, high density uranium compounds and {gamma}-stabilized uranium alloy powder were identified. In Metallic Fuels Division of BARC, R and D efforts are on to develop these high density uranium base alloys. This paper describes the preparation flow sheet for different compositions of Uranium and molybdenum alloys by an innovative powder processing route with uranium and molybdenum metal powders as starting materials. The same composition of U-Mo alloys were also fabricated by conventional method i.e. ingot metallurgy route. The U-Mo alloys prepared by both the methods were then characterized by XRD for phase analysis. The photomicrographs of alloys with different compositions prepared by powder metallurgy and ingot metallurgy routes are also included in the paper. The paper also covers the comparison of properties of the alloys prepared by powder metallurgy and ingot metallurgy routes.

  10. Study of the pyrophoric characteristics of uranium-iron alloys; Etude du caractere pyrophorique des alliages uranium fer

    Energy Technology Data Exchange (ETDEWEB)

    Duplessis, X

    2000-02-23

    The objective of the study is to understand the pyrophoric characteristics of uranium-iron alloys. In order to carry out this research we have elected to use uranium-iron alloy powder with granules of 200 {mu}m and 1000 {mu}m diameter with 4%, 10.8% and 14% iron content. The experiments were performed on small samples of few milligrams and on larger quantities of few hundred grams. The main conclusions obtained are the followings: -The reaction start at 453 K (180 deg. C) and the ignition at 543 K (270 deg. C) - The influence of the specific area seems more important than the iron concentration in the alloys - When the alloy ignites, the fire spreads quickly and the alloy rapidly consumes. (author)

  11. Alloys of uranium and aluminium with low aluminium content; Alliages uranium-aluminium a faible teneur en aluminium

    Energy Technology Data Exchange (ETDEWEB)

    Cabane, G; Englander, M; Lehmann, J [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1955-07-01

    Uranium, as obtained after spinning in phase {gamma}, presents an heterogeneous structure with large size grains. The anisotropic structure of the metal leads to an important buckling and surface distortion of the fuel slug which is incompatible with its tubular cladding for nuclear fuel uses. Different treatments have been made to obtain an isotropic structure presenting high thermal stability (laminating, hammering and spinning in phase {alpha}) without success. Alloys of uranium and aluminium with low aluminium content present important advantage in respect of non allied uranium. The introduction of aluminium in the form of intermetallic compound (UAl{sub 2}) gives a better resistance to thermal fatigue. Alloys obtained from raw casting present an improved buckling and surface distortion in respect of pure uranium. This improvement is obtained with uranium containing between 0,15 and 0,5 % of aluminium. An even more improvement in thermal stability is obtained by thermal treatments of these alloys. These new characteristics are explained by the fine dispersion of the UAl{sub 2} particles in uranium. The results after treatments obtained from an alloy slug containing 0,4 % of aluminium show no buckling or surface distortion and no elongation. (M.P.)

  12. X-ray diffraction study of reversible deformation mechanisms in the aged uranium-6.5 niobium alloy

    International Nuclear Information System (INIS)

    Carpenter, D.A.

    1985-01-01

    The x-ray diffraction (XRD) data from 200 0 C/2h-aged uranium-6.5 wt % niobium (U-6.5Nb) alloys, taken under stress as a function of strain, revealed a gamma-zero (γ 0 )→ alpha prime-prime (α'') thermoelastic martensitic phase transformation. It was concluded that the primary reversible deformation modes consisted of the movement of γ 0 /α'' interphase interfaces and α'' intervariant interfaces. Specimen elasticity at low strains was associated with the retreat of interphase interfaces. At higher strains, interphase interfaces did not recover significantly on unloading, and elasticity was due primarily to the retreat of α'' intervariant interfaces

  13. Time-dependent leak behavior of flawed Alloy 600 tube specimens at constant pressure

    Energy Technology Data Exchange (ETDEWEB)

    Bahn, Chi Bum, E-mail: bahn@anl.gov [Argonne National Laboratory, Argonne, IL 60439 (United States); Majumdar, Saurin [Argonne National Laboratory, Argonne, IL 60439 (United States); Harris, Charles [United States Nuclear Regulatory Commission, Rockville, MD 20852 (United States)

    2011-10-15

    Leak rate testing has been performed using Alloy 600 tube specimens with throughwall flaws. Some specimens have shown time-dependent leak behavior at constant pressure conditions. Fractographic characterization was performed to identify the time-dependent crack growth mechanism. The fracture surface of the specimens showed the typical features of ductile fracture, as well as the distinct crystallographic facets, typical of fatigue crack growth at low {Delta}K level. Structural vibration appears to have been caused by the oscillation of pressure, induced by a high-pressure pump used in a test facility, and by the water jet/tube structure interaction. Analyses of the leak behaviors and crack growth indicated that both the high-pressure pump and the water jet could significantly contribute to fatigue crack growth. To determine whether the fatigue crack growth during the leak testing can occur solely by the water jet effect, leak rate tests at constant pressure without the high-pressure pump need to be performed. - Highlights: > Leak rate of flawed Alloy 600 tubing increased at constant pressure condition. > Fractography revealed two cases: ductile tearing and crystallographic facets. > Crystallographic facets are typical features of fatigue crack growth at low {Delta}K. > Fatigue source could be water jet-induced vibration and/or high-pressure pump pulsation.

  14. Design of high density gamma-phase uranium alloys for LEU dispersion fuel applications

    International Nuclear Information System (INIS)

    Hofman, Gerard L.; Meyer, Mitchell K.; Ray, Allison E.

    1998-01-01

    Uranium alloys are candidates for the fuel phase in aluminium matrix dispersion fuels requiring high uranium loading. Certain uranium alloys have been shown to have good irradiation performance at intermediate burnup. previous studies have shown that acceptable fission gas swelling behavior and fuel-aluminium interaction is possible only if the fuel alloy can be maintained in the high temperature body-centered-cubic γ-phase during fabrication and irradiation, at temperatures at which αU is the equilibrium phase. transition metals in Groups V through VIII are known to allow metastable retention of the gamma phase below the equilibrium isotherm. These metals have varying degrees of effectiveness in stabilizing the gamma phase. Certain alloys are metastable for very long times at the relatively low fuel temperatures seen in research operation. In this paper, the existing data on the gamma stability of binary and ternary uranium alloys is analysed. The mechanism and kinetics of decomposition of the gamma phase are assessed with the help of metal alloy theory. Alloys with the highest possible uranium content, good gamma-phase stability, and good neutronic performance are identified for further metallurgical studies and irradiation tests. Results from theory will be compared with experimentally generated data. (author)

  15. Investigation of smooth specimen scc test procedures; variations in environment, specimen size, stressing frame, and stress state. [for high strength aluminum alloys

    Science.gov (United States)

    Lifka, B. W.; Sprowls, D. O.; Kelsey, R. A.

    1975-01-01

    The variables studied in the stress-corrosion cracking performance of high strength aluminum alloys were: (1) corrosiveness of the environment, (2) specimen size and stiffness of the stressing system, (3) interpretation of transgranular cracking, and (4) interaction of the state of stress and specimen orientation in a product with an anisotropic grain structure. It was shown that the probability of failure and time to fracture for a specimen loaded in direct tension are influenced by corrosion pattern, the stressing assembly stiffness, and the notch tensile strength of the alloy. Results demonstrate that the combination of a normal tension stress and a shear stress acting on the plane of maximum susceptibility in a product with a highly directional grain cause the greatest tendency for stress-corrosion cracking.

  16. Atmospheric corrosion of uranium-carbon alloys; Corrosion atmospherique des alliages uranium-carbone

    Energy Technology Data Exchange (ETDEWEB)

    Rousset, P; Accary, A [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1965-07-01

    The authors study the corrosion of uranium-carbon alloys having compositions close to that of the mono-carbide; they show that the extent of the observed corrosion effects increases with the water vapour content of the surrounding gas and they conclude that the atmospheric corrosion of these alloys is due essentially to the humidity of the air, the effect of the oxygen being very slight at room temperature. They show that the optimum conditions for preserving U-C alloys are either a vacuum or a perfectly dry argon atmosphere. The authors have also established that the type of corrosion involved is a corrosion which 'cracks under stress' and is transgranular (it can also be intergranular in the case of sub-stoichiometric alloys). They propose, finally, two hypotheses for explaining this mechanism, one of which is illustrated by the existence, at the fissure interface, of corrosion products which can play the role of 'corners' in the mono-carbide grains. (authors) [French] Les auteurs etudient la corrosion des alliages uranium-carbone de composition voisine du monocarbure; ils montrent que l'importance des effets de la corrosion observee augmente avec la teneur en vapeur d'eau du milieu gazeux ambiant et concluent que la corrosion atmospherique de ces alliages est due essentiellement a l'humidite de l'air, l'action de l'oxygene de l'air etant tres faible a la temperature ambiante. Ils indiquent que les conditions optimales de conservation des alliages U-C sont le vide ou une atmosphere d'argon parfaitement desseches. D'autre part, les auteurs etablissent que le type de corrosion mis en jeu est une corrosion 'fissurante sous contrainte', transgranulaire (pouvant egalement etre intergranulaire dans le cas d'alliages sous-stoechiometriques). Ils proposent enfin deux hypotheses pour rendre compte de ce mecanisme, dont l'une est illustree par la mise en evidence, a l'interface des fissures, de produits de corrosion pouvant jouer le role de 'coins' dans les grains de

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

    International Nuclear Information System (INIS)

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

    1975-01-01

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

  18. Analysis of uranium and of some of its compounds and alloys. Copper spectrophotometric determination

    International Nuclear Information System (INIS)

    Copper determination in uranium, uranium oxides (UO 2 , UO 3 , U 3 O 8 ), ammonium diuranate, U-Al-Fe alloy (700 ppm Al and 300 ppm Fe) and U-Mo alloy (1.1 percent Mo) by acid dissolution reduction of copper by hydroxylamine hydrochloride and formation of a complex with diquinolyle-2,2' amyl alcohol (pH value 6 to 7) and spectrophotometry at 550 nm. The method is applicable for copper content between 5 to 40 ppm in respect of uranium contained in the material [fr

  19. Thermal Analysis of Pure Uranium Metal, UMo and UMoSi Alloys Using a Differential Thermal Analyzer

    International Nuclear Information System (INIS)

    Yanlinastuti; Sutri Indaryati; Rahmiati

    2010-01-01

    Thermal analysis of pure uranium metal, U-7%Mo and U-7%Mo-1%Si alloys have been done using a Differential Thermal Analyzer (DTA). The experiments are conducted in order to measure the thermal stability, thermochemical properties of elevated temperature and enthalpy of the specimens. From the analysis results it is showed that uranium metal will transform from α to β phases at temperature of 667.16°C and enthalpy of 2.3034 cal/g and from β to γ phases at temperature of 773.05 °C and enthalpy of 2.8725 cal/g and start melting at temperature of 1125.26 °C and enthalpy of 2.1316 cal/g. The U-7%Mo shows its thermal stability up to temperature of 650 °C and its thermal changes at temperature of 673.75 °C indicated by the formation of an endothermic peak and enthalpy of 0.0257 cal/g. The U-7%Mo-1%Si alloys shows its thermal stability up to temperature of 550 °C and its thermal changes at temperature of 574.18 °C indicated by the formation of an endothermic peak and enthalpy of 0.613 cal/g. From the three specimens it is showed that they have a good thermal stability at temperature up to 550 °C. (author)

  20. Stress Wave Attenuation in Aluminum Alloy and Mild Steel Specimens Under SHPB Tensile Testing

    Science.gov (United States)

    Pothnis, J. R.; Ravikumar, G.; Arya, H.; Yerramalli, Chandra S.; Naik, N. K.

    2018-02-01

    Investigations on the effect of intensity of incident pressure wave applied through the striker bar on the specimen force histories and stress wave attenuation during split Hopkinson pressure bar (SHPB) tensile testing are presented. Details of the tensile SHPB along with Lagrangian x- t diagram of the setup are included. Studies were carried out on aluminum alloy 7075 T651 and IS 2062 mild steel. While testing specimens using the tensile SHPB setup, it was observed that the force calculated from the transmitter bar strain gauge was smaller than the force obtained from the incident bar strain gauge. This mismatch between the forces in the incident bar and the transmitter bar is explained on the basis of stress wave attenuation in the specimens. A methodology to obtain force histories using the strain gauges on the specimen during SHPB tensile testing is also presented. Further, scanning electron microscope images and photomicrographs are given. Correlation between the microstructure and mechanical properties is explained. Further, uncertainty analysis was conducted to ascertain the accuracy of the results.

  1. Macro and Microscopic Investigation on Fracture Specimen of Alloy 617 Base Metal and Weldment in Low Cycle Fatigue Regime

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Seon Jin; Dewa, Rando Tungga [Pukyung National Univ., Busan (Korea, Republic of); Kim, Won Gon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-06-15

    This paper investigates macro- and microscopic fractography performed on fracture specimens from low cycle fatigue (LCF) testings through an Alloy 617 base metal and weldments. The weldment specimens were taken from gas tungsten arc welding (GTAW) pad of Alloy 617. The aim of the present study is to investigate the macro- and microscopic aspects of the low cycle fatigue fracture mode and mechanism of Alloy 617 base metal and GTAWed weldment specimens. Fully axial total strain controlled fatigue tests were conducted at room temperature with total strain ranges of 0.6, 0.9, 1.2 and 1.5%. Macroscopic fracture surfaces of Alloy 617 base metal specimens showed a flat type normal to the fatigue loading direction, whereas the GTAWed weldment specimens were of a shear/star type. The fracture surfaces of both the base metal and weldment specimens revealed obvious fatigue striations at the crack propagation regime. In addition, the fatigue crack mechanism of the base metal showed a transgranular normal to fatigue loading direction; however, the GTAWed weldment specimens showed a transgranular at approximately 45° to the fatigue loading direction.

  2. Macro and Microscopic Investigation on Fracture Specimen of Alloy 617 Base Metal and Weldment in Low Cycle Fatigue Regime

    International Nuclear Information System (INIS)

    Kim, Seon Jin; Dewa, Rando Tungga; Kim, Won Gon

    2016-01-01

    This paper investigates macro- and microscopic fractography performed on fracture specimens from low cycle fatigue (LCF) testings through an Alloy 617 base metal and weldments. The weldment specimens were taken from gas tungsten arc welding (GTAW) pad of Alloy 617. The aim of the present study is to investigate the macro- and microscopic aspects of the low cycle fatigue fracture mode and mechanism of Alloy 617 base metal and GTAWed weldment specimens. Fully axial total strain controlled fatigue tests were conducted at room temperature with total strain ranges of 0.6, 0.9, 1.2 and 1.5%. Macroscopic fracture surfaces of Alloy 617 base metal specimens showed a flat type normal to the fatigue loading direction, whereas the GTAWed weldment specimens were of a shear/star type. The fracture surfaces of both the base metal and weldment specimens revealed obvious fatigue striations at the crack propagation regime. In addition, the fatigue crack mechanism of the base metal showed a transgranular normal to fatigue loading direction; however, the GTAWed weldment specimens showed a transgranular at approximately 45° to the fatigue loading direction

  3. Postirradiation examination of high-density uranium alloy dispersion fuels

    International Nuclear Information System (INIS)

    Hayes, S.L.; Meyer, M.K.; Hofman, G.L.; Strain, R.V.

    1998-01-01

    Two irradiation test vehicles, designated RERTR-2, were inserted into the Advanced Test reactor in Idaho in August 1997. These tests were designed to obtain irradiation performance information on a variety of potential new, high-density uranium alloy dispersion fuels, including U-10Mo, U-8Mo, U-6Mo, U-4Mo, U-9Nb-3Zr, U-6Nb-4Zr, U-5Nb-3Zr, U-6Mo-1Pt, U-6Mo-0.6Ru and U-10Mo-0.05Sn: the intermetallic compounds U 2 Mo and U-10Mo-0.-5Sn; the intermetallic compounds U 2 Mo and U 3 Si 2 were also included in the fuel test matrix. These fuels are included in the experiments as microplates (76 mm x 22 mm x 1.3mm outer dimensions) with a nominal fuel volume loading of 25% and irradiated at relatively low temperature (∼100 deg C). RERTR-1 and RERTR-2 were discharged from the reactor in November 1997 and July 1998, respectively at calculated peak fuel burnups of 45 and 71 at %-U 235 Both experiments are currently under examination at the Alpha Gamma Hot Cell Facility at Argonne National Laboratory in Chicago. This paper presents the postirradiation examination results available to date from these experiments. (author)

  4. 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

  5. Annex 4 - Task 08/13 final report, Producing the binary uranium alloys with alloying components Al, Mo, Zr, Nb, and B

    International Nuclear Information System (INIS)

    Lazarevic, Dj.

    1961-01-01

    Due to reactivity of uranium in contact with the gasses O 2 , N 2 , H 2 , especially under higher temperatures uranium processing is always done in vacuum or inert gas. Melting, alloying and casting is done in high vacuum stoves. This report reviews the type of furnaces and includes detailed description of the electric furnace for producing uranium alloys which is available in the Institute

  6. Impact strength of the uranium-6 weight percent niobium alloy between -1980 and +2000C

    International Nuclear Information System (INIS)

    Anderson, R.C.

    1981-09-01

    A study was conducted to determine if a ductile-to-brittle transition wxisted for the uranium-6 wt % niobium (U-6Nb) alloy. Standard V-notched Charpy bars were made from both solution-quenched and solution-quenched and aged U-6Nb alloy and were tested between -198 0 and +200 0 C. It was found that a sharp ductile-brittle transition does not exist for the alloy. A linear relationship existed between test temperature and impact strength, and the alloy retained a significant amount of impact strength even at very low temperatures. 9 figures

  7. Nuclear reactor fuel structure containing uranium alloy wires embedded in a metallic matrix plate

    Science.gov (United States)

    Travelli, Armando

    1988-01-01

    A flat or curved plate structure, to be used as fuel in a nuclear reactor, comprises elongated fissionable wires or strips embedded in a metallic continuous non-fissionable matrix plate. The wires or strips are made predominantly of a malleable uranium alloy, such as uranium silicide, uranium gallide or uranium germanide. The matrix plate is made predominantly of aluminum or an aluminum alloy. The wires or strips are located in a single row at the midsurface of the plate, parallel with one another and with the length dimension of the plate. The wires or strips are separated from each other, and from the surface of the plate, by sufficient thicknesses of matrix material, to provide structural integrity and effective fission product retention, under neutron irradiation. This construction makes it safely feasible to provide a high uranium density, so that the uranium enrichment with uranium 235 may be reduced below about 20%, to deter the reprocessing of the uranium for use in nuclear weapons.

  8. Study of the quenching and subsequent return to room temperature of uranium-chromium, uranium-iron, and uranium-molybdenum alloys containing only small amounts of the alloying element

    International Nuclear Information System (INIS)

    Delaplace, J.

    1960-09-01

    By means of an apparatus which makes possible thermal pre-treatments in vacuo, quenching carried out in a high purity argon atmosphere, and simultaneous recording of time temperature cooling and thermal contraction curves, the author has examined the transformations which occur in uranium-chromium, uranium-iron and uranium-molybdenum alloys during their quenching and subsequent return to room temperature. For uranium-chromium and uranium-iron alloys, the temperature at which the γ → β transformation starts varies very little with the rate of cooling. For uranium-molybdenum alloys containing 2,8 atom per cent of Mo, this temperature is lowered by 120 deg. C for a cooling rate of 500 deg. C/mn. The temperature at which the β → α transformation starts is lowered by 170 deg. C for a cooling rate of 500 deg. C/mn in the case of uranium-chromium alloy containing 0,37 atom per cent of Cr. The temperature is little affected in the case of uranium-iron alloys. The addition of chromium or iron makes it possible to conserve the form β at ordinary temperatures after quenching from the β and γ regions. The β phase is particularly unstable and changes into needles of the α form even at room temperatures according to an autocatalytic transformation law similar to the austenitic-martensitic transformation law in the case of iron. The β phase obtained by quenching from the β phase region is more stable than that obtained by quenching from the γ region. Chromium is a more effective stabiliser of the β phase than is iron. Unfortunately it causes serious surface cracking. The β → α transformation in uranium-chromium alloys has been followed at room temperature by means of micro-cinematography. The author has not observed the direct γ → α transformation in uranium-molybdenum alloys containing 2,8 per cent of molybdenum even for cooling rates of up to 2000 deg. C/s. He has however observed the formation of several martensitic structures. (author) [fr

  9. Solubility of hydrogen and deuterium in bcc-uranium-titanium alloys

    International Nuclear Information System (INIS)

    Powell, G.L.; Kirkpatrick, J.R.

    1996-01-01

    For the bcc-U-Ti alloy system, H and D solubility measurements have been made on 12 alloy specimens ranging in composition from pure U to pure Ti and temperature range bounded by 900 K to 1,500 K. The results are described by a model within a standard error of 3%

  10. 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)

  11. Gaseous oxygen and hydrogen embrittlements of the uranium-10 weight % molybdenum alloy

    International Nuclear Information System (INIS)

    Corcos, Jean.

    1979-07-01

    The stress corrosion of an Uranium-10 weight % Molybdenum alloy in high purity gaseous oxygen and hydrogen was studied. Tests were performed with fracture-mechanic specimens, fatigue precracked and carried out in tension with a constant sustained load. The experimental procedure enabled to determine the S.C. morphology during the test, and its kinetics. Tests in gaseous oxygen were performed with p02=0.15 MPa from 0 0 C to 100 0 C, and at 20 0 C for p02=0.15, 0.15.10 -2 and 0.15.10 -4 MPa. Two kinetic laws are proposed. Cracking is transgranular with a quasi-clivage type, and occurs on the (1 1 1) planes of the matrix. Tests in gaseous hydrogen were performed with pH2=0.15 MPa from - 50 0 C to + 135 0 C; for all the tests, even those under no exterior load, there is a failure by S.C. and macroscopic hydruration occurs. We propose a kinetic law, which may display that the hydruration phenomenon rules the S.C. propagation. We have performed the identification of the hydride, as well as the study of the precipitation. These phenomena don't occur with pH2=0.15.10 -2 MPa. The embrittlement is thought to be due to a formation-failure cycle of an hydride precipitate at the crack tip [fr

  12. Obtention of uranium-molybdenum alloy ingots technique to avoid carbon contamination

    Energy Technology Data Exchange (ETDEWEB)

    Pedrosa, Tercio A.; Paula, Joao Bosco de; Reis, Sergio C.; Brina, Jose Giovanni M.; Faeda, Kelly Cristina M.; Ferraz, Wilmar B., E-mail: tap@cdtn.b, E-mail: jbp@cdtn.b, E-mail: jgmb@cdtn.b, E-mail: ferrazw@cdtn.b [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

    2011-07-01

    The replacement of high enriched uranium (U{sup 235} > 85 wt%) by low enriched uranium (U{sup 235} < 20wt%) nuclear fuels in research and test reactors is being implemented as an initiative of the Reduced Enrichment for Research and Test Reactors (RERTR) program, conceived in the USA since mid-70s, in order to avoid nuclear weapons proliferation. Such replacement implies in the use of compounds or alloys with higher uranium densities. Among the several uranium alloys investigated since then, U-Mo presents great application potential due to its physical properties and good behavior during irradiation, which makes it an important option as a nuclear fuel material for the Brazilian Multipurpose Reactor - RMB. The development of the plate-type nuclear fuel based on U-Mo alloy is being performed at the Nuclear Technology Development Centre (CDTN) and also at IPEN. The carbon contamination of the alloy is one of the great concerns during the melting process. It was observed that U-Mo alloy is more critical considering carbon contamination when using graphite crucibles. Alternative melting technique was implemented at CDTN in order to avoid carbon contamination from graphite crucible using Yttria stabilized ZrO{sub 2} crucibles. Ingots with low carbon content and good internal quality were obtained. (author)

  13. Microstructural investigation of as-cast uranium rich U–Zr alloys

    Energy Technology Data Exchange (ETDEWEB)

    Zhang, Yuting, E-mail: zhangyuting@caep.cn [Science and Technology on Surface Physics and Chemistry Laboratory, Jiangyou 621908, Sichuan (China); School of Nuclear Science and Technology, National Synchrotron Radiation Laboratory, University of Science and Technology of China, Hefei 230029, Anhui (China); Wang, Xin [Science and Technology on Surface Physics and Chemistry Laboratory, Jiangyou 621908, Sichuan (China); Zeng, Gang [Institute of Materials, China Academy of Engineering Physics, Jiangyou 621908, Sichuan (China); Wang, Hui [Science and Technology on Surface Physics and Chemistry Laboratory, Jiangyou 621908, Sichuan (China); Jia, Jianping [Institute of Materials, China Academy of Engineering Physics, Jiangyou 621908, Sichuan (China); Sheng, Liusi [School of Nuclear Science and Technology, National Synchrotron Radiation Laboratory, University of Science and Technology of China, Hefei 230029, Anhui (China); Zhang, Pengcheng, E-mail: zpc113@sohu.com [Science and Technology on Surface Physics and Chemistry Laboratory, Jiangyou 621908, Sichuan (China)

    2016-04-01

    The present study evaluates the microstructure in as-cast uranium rich U–Zr alloys, an important subsystem of U–Pu–Zr ternary metallic nuclear reactor fuel, as a function of the Zr content, from 2wt.% to 15wt.%Zr. It has been previously suggested that the unique intermetallic compound δ phase in U–Zr alloys is only present in as-cast U–Zr alloys with a Zr content exceeding 10wt.%Zr. However, our analysis of transmission electron microscopy (TEM) data shows that the δ phase is common to all as-cast alloys studied in this work. Furthermore, specific coherent orientation relationship is found between the α and δ phases, consistent with previous findings, and a third variant is discovered in this paper. - Highlights: • Initially, lattice parameter of as-cast U–Zr alloys decrease with the increasing Zr content, and then increase. • XRD data show the presence of δ-UZr{sub 2} phase in as-cast U–Zr alloys with a Zr content of more than 8wt.% Zr. • Finding δ-UZr{sub 2} phase exists in all as-cast uranium rich U–Zr alloys, even for alloys with a lean Zr content. • Three kinds of preferential orientations of the δ phase grow.

  14. A review of the environmental behavior of uranium derived from depleted uranium alloy penetrators

    Energy Technology Data Exchange (ETDEWEB)

    Erikson, R.L.; Hostetler, C.J.; Divine, J.R.; Price, K.R.

    1990-01-01

    The use of depleted uranium (DU) penetrators as armor-piercing projectiles in the field results in the release of uranium into the environment. Elevated levels of uranium in the environment are of concern because of radioactivity and chemical toxicity. In addition to the direct contamination of the soil with uranium, the penetrators will also chemically react with rainwater and surface water. Uranium may be oxidized and leached into surface water or groundwater and may subsequently be transported. In this report, we review some of the factors affecting the oxidation of the DU metal and the factors influencing the leaching and mobility of uranium through surface water and groundwater pathways, and the uptake of uranium by plants growing in contaminated soils. 29 refs., 10 figs., 3 tabs.

  15. 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)

  16. Low alloy additions of iron, silicon, and aluminum to uranium: a literature survey

    International Nuclear Information System (INIS)

    Ludwig, R.L.

    1980-01-01

    A survey of the literature has been made on the experimental results of small additions of iron, silicon, and aluminum to uranium. Information is also included on the constitution, mechanical properties, heat treatment, and deformation of various binary and ternary alloys. 42 references, 24 figures, 13 tables

  17. Preparation, heat treatment, and mechanical properties of the uranium-5 weight percent chromium eutectic alloy

    International Nuclear Information System (INIS)

    Townsend, A.B.

    1980-10-01

    The eutectic alloy of uranium-5 wt % chromium (U-5Cr) was prepared from high-purity materials and cast into 1-in.-thick ingots. This material was given several simple heat treatments, the mechanical properties of these heat-treated samples were determined; and the microstructure was examined. Some data on the melting point and transformation temperatures were obtained

  18. Stress field determination in an alloy 600 stress corrosion crack specimen

    International Nuclear Information System (INIS)

    Rassineux, B.; Labbe, T.

    1995-05-01

    In the context of EDF studies on stress corrosion cracking rates in the Alloy 600 steam generators tubes, we studied the influence of strain hardened surface layers on the different stages of cracking for a tensile smooth specimen (TLT). The stress field was notably assessed to try and explain the slow/rapid-propagation change observed beyond the strain hardened layers. The main difficulty is to simulate in a finite element model the inner and outer surfaces of these strain hardened layers, produced by the final manufacturing stages of SG tubes which have not been heat treated. In the model, the strain hardening is introduced by simulating a multi-layer material. Residual stresses are simulated by an equivalent fictitious thermomechanical calculation, realigned with respect to X-ray measurements. The strain hardening introduction method was validated by an analytical calculation giving identical results. Stress field evolution induced by specimen tensile loading were studied using an elastoplastic 2D finite element calculations performed with the Aster Code. The stress profile obtained after load at 660 MPa shows no stress discontinuity at the boundary between the strain hardened layer and the rest of the tube. So we propose that a complementary calculation be performed, taking into account the multi-cracked state of the strain hardened zones by means of a damage variable. In fact, this state could induce stress redistribution in the un-cracked area, which would perhaps provide an explanation of the crack-ground rate change beyond the strain hardened zone. The calculations also evidence the harmful effects of plastic strains on a strain hardened layer due to the initial state of the tube (not heat-treated), to grit blasting or to shot peening. The initial compressive stress condition of this surface layer becomes, after plastic strain, a tensile stress condition. These results are confirmed by laboratory test. (author). 10 refs., 18 figs., 9 tabs., 2 appends

  19. Determination of crystalline texture in aluminium - uranium alloys by neutron diffraction

    International Nuclear Information System (INIS)

    Azevedo, A.M.V. de.

    1978-01-01

    Textures of hot-rolled aluminum-uranium alloys and of aluminum were determined by neutron diffraction. Sheets of alloys containing 8.0, 21.5 and 23.7 wt pct U, as well as pure aluminum, were obtained in a stepped rolling process, 15% reduction each step, 75% total reduction. During the rolling the temperature was 600 0 C. Alloys with low uranium contents are two phase systems in which an intermetallic compound UAl 4 , orthorhombic, is dispersed in a pure aluminum matrix. The addition of a few percent of Si in such alloys leads to the formation of UAl 3 , simple cubic, instead of UAl 4 . The Al -- 23.7 wt pct U alloy was prepared with 2,2 wt pct of Si. The results indicate that the texture of the matrix is more dependent on the uranium concentration than on the texture of the intermetallic phases. An improvement in the technique applied to texture measurements by using a sample fully bathed in the neutron beam is also presented. The method takes advantage of the low neutron absorption of the studied materials as well as of the neglibible variation in the multiple scattering which occurs in a conveniently shaped sample having a weakly developed texture. (Author) [pt

  20. The detection of tightly closed flaws by nondestructive testing (NDT) methods. [fatigue crack formation in aluminum alloy test specimens

    Science.gov (United States)

    Rummel, W. D.; Rathke, R. A.; Todd, P. H., Jr.; Mullen, S. J.

    1975-01-01

    Liquid penetrant, ultrasonic, eddy current and X-radiographic techniques were optimized and applied to the evaluation of 2219-T87 aluminum alloy test specimens in integrally stiffened panel, and weld panel configurations. Fatigue cracks in integrally stiffened panels, lack-of-fusion in weld panels, and fatigue cracks in weld panels were the flaw types used for evaluation. A 2319 aluminum alloy weld filler rod was used for all welding to produce the test specimens. Forty seven integrally stiffened panels containing a total of 146 fatigue cracks, ninety three lack-of-penetration (LOP) specimens containing a total of 239 LOP flaws, and one-hundred seventeen welded specimens containing a total of 293 fatigue cracks were evaluated. Nondestructive test detection reliability enhancement was evaluated during separate inspection sequences in the specimens in the 'as-machined or as-welded', post etched and post proof loaded conditions. Results of the nondestructive test evaluations were compared to the actual flaw size obtained by measurement of the fracture specimens after completing all inspection sequences. Inspection data were then analyzed to provide a statistical basis for determining the flaw detection reliability.

  1. Study of uranium-plutonium alloys containing from 0 to 20 peri cent of plutonium (1963)

    International Nuclear Information System (INIS)

    Paruz, H.

    1963-05-01

    The work is carried out on U-Pu alloys in the region of the solid solution uranium alpha and in the two-phase region uranium alpha + the zeta phase. The results obtained concern mainly the influence of the addition of plutonium on the physical properties of the uranium (changes in the crystalline parameters, the density, the hardness) in the region of solid solution uranium alpha. In view of the discrepancies between various published results as far as the equilibrium diagram for the system U-Pu is concerned, an attempt was made to verify the extent of the different regions of the phase diagram, in particular the two phased-region. Examinations carried out on samples after various thermal treatments (in particular quenching from the epsilon phase and prolonged annealings, as well as a slow cooling from the epsilon phase) confirm the results obtained at Los Alamos and Harwell. (author) [fr

  2. X-ray topography of uranium alloys; Topographie aux rayons X d'alliages d'uranium

    Energy Technology Data Exchange (ETDEWEB)

    Le Naour, L [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1968-07-01

    A description of the structure of uranium alloys has been made using the data obtained by X-ray diffraction techniques derived from the Berg-Barrette method. In the first.stage the use of a monochromatic beam of X-rays having a very low divergence makes it possible to obtain very reproducible and exact numerical data concerning the grain and sub-grain sizes, and also the distribution of the sizes. It is thereby possible to detect any disorientation greater than 30 seconds of arc.The results obtained have been completed using a variable incidence device which- gives simultaneously an overall picture of a grain and an idea of the importance of internal disorientations; a more rigorous measurement of this latter parameter is then deduced from the Debye-Scherrer diagrams obtained using a fine-focus equipment. Observations are carried out on various one-phase or two phase uranium alloys which are compared successively to technical and to high-purity uranium. It is shown that the use of X-ray topographies, although limited in certain respects, allows a quantitative characterization of the structure. (author) [French] Une description des structures d'alliages d'uranium a ete faite a partir des donnees fournies par des techniques de diffraction de rayons X derivees de la methode de BERG--BARRETT. Dans une premiere etape, l'utilisation d'un faisceau de rayons X monochromatique et de tres faible divergence permet d'obtenir des donnees numeriques precises et tres reproductibles, relatives aux dimensions des grains, des sous-grains et a la distribution de ces grandeurs. Toute desorientation superieure a 30 secondes d'arc peut ainsi etre decelee. Les resultats obtenus ont ete completes en utilisant un montage a incidence variable, qui fournit simultanement l'image globale d'un grain et l'ordre de grandeur des desorientations internes; une mesure plus rigoureuse de ce dernier parametre se deduit ensuite de diagrammes DEBYE SHERRER realises avec un montage a foyer fin. Des

  3. Obtention of uranium-molybdenum alloy ingots microstructure and phase characterization

    Energy Technology Data Exchange (ETDEWEB)

    Pedrosa, Tercio A.; Braga, Daniel M.; Paula, Joao Bosco de; Brina, Jose Giovanni M.; Ferraz, Wilmar B., E-mail: tap@cdtn.b, E-mail: bragadm@cdtn.b, E-mail: jbp@cdtn.b, E-mail: jgmb@cdtn.b, E-mail: ferrazw@cdtn.b [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

    2011-07-01

    The replacement of high enriched uranium (U-{sup 235} > 85 wt%) by low enriched uranium (U-{sup 235} < 20 wt%) nuclear fuels in research and test reactors is being implemented as an initiative of the Reduced Enrichment for Research and Test Reactors (RERTR) program, conceived in the USA since mid-70s, in order to avoid nuclear weapons proliferation. Such replacement implies in the use of compounds or alloys with higher uranium densities. Several uranium alloys that fill this requirement has been investigated since then. Among these alloys, U-Mo presents great application potential due to its physical properties and good behavior during irradiation, which makes it an important option as a nuclear fuel material for the Brazilian Multipurpose Reactor - RMB. The development of the plate-type nuclear fuel based on U-Mo alloys is being performed at the Nuclear Technology Development Centre (CDTN) and also at the Institute of Energetic and Nuclear Research - IPEN. U-{sup 10}Mo ingots were melted in an induction furnace with protective argon atmosphere. The microstructure of the ingots were characterized through optical and scanning electronic microscopy in the as cast and heat treated conditions. Energy Dispersive Spectrometry and X-Ray Diffraction were used as characterization techniques for elemental analysis and phases determination. It was confirmed the presence of metastable gamma-phase in the as cast condition, surrounded by hypereutectoid alpha-phase (uranium-rich phase), as well as a pearlite-like constituent, composed by alternated lamellas of U{sub 2}Mo compound and alpha-phase, in the heat treated condition. (author)

  4. Vapor corrosion of aluminum cladding alloys and aluminum-uranium fuel materials in storage environments

    International Nuclear Information System (INIS)

    Lam, P.; Sindelar, R.L.; Peacock, H.B. Jr.

    1997-04-01

    An experimental investigation of the effects of vapor environments on the corrosion of aluminum spent nuclear fuel (A1 SNF) has been performed. Aluminum cladding alloys and aluminum-uranium fuel alloys have been exposed to environments of air/water vapor/ionizing radiation and characterized for applications to degradation mode analysis for interim dry and repository storage systems. Models have been developed to allow predictions of the corrosion response under conditions of unlimited corrodant species. Threshold levels of water vapor under which corrosion does not occur have been identified through tests under conditions of limited corrodant species. Coupons of aluminum 1100, 5052, and 6061, the US equivalent of cladding alloys used to manufacture foreign research reactor fuels, and several aluminum-uranium alloys (aluminum-10, 18, and 33 wt% uranium) were exposed to various controlled vapor environments in air within the following ranges of conditions: Temperature -- 80 to 200 C; Relative Humidity -- 0 to 100% using atmospheric condensate water and using added nitric acid to simulate radiolysis effects; and Gamma Radiation -- none and 1.8 x 10 6 R/hr. The results of this work are part of the body of information needed for understanding the degradation of the A1 SNF waste form in a direct disposal system in the federal repository. It will provide the basis for data input to the ongoing performance assessment and criticality safety analyses. Additional testing of uranium-aluminum fuel materials at uranium contents typical of high enriched and low enriched fuels is being initiated to provide the data needed for the development of empirical models

  5. Stress corrosion cracking of uranium--niobium alloys

    International Nuclear Information System (INIS)

    Magnani, N.J.

    1978-03-01

    The stress corrosion cracking behavior of U-2 1 / 4 , 4 1 / 2 , 6 and 8 wt % Nb alloys was evaluated in laboratory air and in aqueous Cl - solutions. Thresholds for crack propagation were obtained in these environments. The data showed that Cl - solutions are more deleterious than air environments. Tests were also conducted in pure gases to identify the species in the air responsible for cracking. These data showed the primary stress corrodent is water vapor for the most reactive alloy, U-2 1 / 4 % Nb, while O 2 is primarily responsible for cracking in the more corrosion resistant alloys, U-6 and 8% Nb. The 4 1 / 2 % alloy was found to be susceptible in both H 2 O and O 2 environments

  6. Dilatometric studies on uranium-zirconium-fissium alloy

    International Nuclear Information System (INIS)

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

    2012-01-01

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

  7. Contribution towards the study of β→α transformation in uranium and its alloys (1962)

    International Nuclear Information System (INIS)

    Aubert, H.

    1962-05-01

    The kinetics of the transformation of uranium alloys containing 0.5 - 0.75 - 1.0 - 1.5 and 3 atoms per cent have been studied. The influence of heat treatment before decomposition has been discussed. The study of the transformation characteristics such as kinetics, residual phases, phenomena connected with the coherence between phases, reversibility below the equilibrium temperature, shows the following mechanisms exhibited during the decomposition of the β phase on lowering the temperature: 1 ) eutectoid, 2) bainitic, 3) martensitic. The study of the TTT diagrams of alloys containing decreasing percentages of chromium indicates that the unalloyed uranium transforms without maintaining the coherence above 600 deg. C, where as at lower temperatures the transformation is mainly martensitic. The various alloying elements can be characterised by their influence on the three TTT curves corresponding to the three possible transformation mechanisms. The ability of the uranium alloys to alpha grain refining during isothermal decomposition or ambient temperature quenching is directly connected with the characteristics of the TTT diagrams and especially to the mode of bainitic transformation. (author) [fr

  8. Electrolytic etching of uranium and of its alloys for examination under ordinary light

    International Nuclear Information System (INIS)

    Bouleau, M.

    1958-12-01

    The author reports a metallographic study of uranium and of some of its alloys (U-Mo with different Mo contents, U-Sn, U-Al) performed by using electrolytic etching. Samples are polished before being etched. Metallographic images are provided and results are briefly stated in terms of presence of grain boundaries, twins, platelets, pitting, metallic and non-metallic inclusions or eutectoid decomposition. The authors notice that, in some alloys with a gamma-stabilized structure, electrolytic etching allows an oxidation under reduced oxygen pressure, and then phase structure to be perfectly revealed

  9. Highlighting micrographic structures of uranium alloys containing 0.5 to 10 per cent wt molybdenum

    International Nuclear Information System (INIS)

    Laniesse, J.; Bouleau, M.

    1959-02-01

    The authors report a study which aimed at determining for different uranium molybdenum alloys and with respect to their molybdenum content a polishing method which allows a relatively simple grain examination in the as-cast condition, an as perfect as possible resolution of eutectic decompositions, and the appropriate conditions to highlight structures (beta-alpha and gamma-alpha martensite transformations, beta phase retention and decomposition, transient structures, eutectoid decomposition, and so on). Alloys differ by their molybdenum content: from 0.5 to 1 per cent wt, 1.5 to 3 per cent wt, 5 to 10 per cent wt

  10. 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)

  11. The fracture mechanism of uranium-niobium alloys near hypoeutectoid composition aged at low temperature

    International Nuclear Information System (INIS)

    Wang Xiaoying; Ren Dapeng; Yang Jianxiong; Jiang Guifen

    2006-01-01

    The microstructures and the crack propagation of uranium-niobium alloys near hypoeutectoid composition aged at temperature 200 degree C for 2 hours during a tension was investigated by means of in situ tension tests using TEM. The results show that the twinning planes inside and between the martensite laths move and merge, and then disintegrate in uranium-niobium alloys with monoclinic α structure during the tension. The crack propagation can be described as follows. Under the tension, the thinning zone which is locally plastically deformed emerges in the front of the crack tip. After the process of nucleation, growth and conjunction, the microvoids connect with the main crack, which results in the fracture. Neither of emission, propagation and movement of dislocation was observed during the tension. (authors)

  12. Fabrication and characterization of uranium-6--niobium alloy plate with improved homogeneity

    International Nuclear Information System (INIS)

    Snyder, W.B.

    1978-01-01

    Chemical inhomogeneities produced during arc melting of uranium--6 weight percent niobium alloy normally persist during fabrication of the ingot to a finished product. An investigation was directed toward producing a more homogeneous product (approx. 13.0-mm plate) by a combination of mechanical working and homogenization. Ingots were cast, forged to various reductions, homogenized under different conditions, and finally rolled to 13.0-mm-thick plate. It was concluded that increased forging reductions prior to homogenization resulted in a more homogeneous plate. Comparison of calculated and experimentally measured niobium concentration profiles indicated that the activation energy for the diffusion of niobium in uranium--niobium alloys may be lower than previously observed

  13. A Model for High-Strain-Rate Deformation of Uranium-Niobium Alloys

    Energy Technology Data Exchange (ETDEWEB)

    F.L.Addessio; Q.H.Zuo; T.A.Mason; L.C.Brinson

    2003-05-01

    A thermodynamic approach is used to develop a framework for modeling uranium-niobium alloys under the conditions of high strain rate. Using this framework, a three-dimensional phenomenological model, which includes nonlinear elasticity (equation of state), phase transformation, crystal reorientation, rate-dependent plasticity, and porosity growth is presented. An implicit numerical technique is used to solve the evolution equations for the material state. Comparisons are made between the model and data for low-strain-rate loading and unloading as well as for heating and cooling experiments. Comparisons of the model and data also are made for low- and high-strain-rate uniaxial stress and uniaxial strain experiments. A uranium-6 weight percent niobium alloy is used in the comparisons of model and experiment.

  14. Computer simulation of quenching uranium-0.75 weight per cent titanium alloy

    International Nuclear Information System (INIS)

    Ludtka, G.M.; Llewellyn, G.H.; Aramayo, G.A.; Siman-Tov, M.; Childs, K.W.

    1986-01-01

    A ''QUENCH SIMULATOR'' has been developed which uses finite difference heat transfer and finite element stress analysis techniques to predict the behavior of a metal during quenching. The actual nonlinear temperature- and microstructure-dependent physical, thermophysical, and mechanical properties are incorporated as input into the computer model as well as the continuous cooling transformation (CCT) behavior and heats of transformation of the alloy. The final output provides the transient temperature distribution, details the final residual profile, predicts and shows where distortion occurs, and maps out the microstructure distribution throughout the entire sample. These data are available in tabulated form, contour plots, or color-coded graphics. This analysis has been demonstrated on simple shapes for unalloyed uranium and the uranium-0.75 weight per titanium alloy which undergoes a martensite transformation and is quench-rate sensitive. The results of this study are discussed in detail in addition to other applications of this analysis approach which is generic in nature

  15. Irradiation performance of uranium-molybdenum alloy dispersion fuels

    International Nuclear Information System (INIS)

    Almeida, Cirila Tacconi de

    2005-01-01

    The U-Mo-Al dispersion fuels of Material Test Reactors (MTR) are analyzed in terms of their irradiation performance. The irradiation performance aspects are associated to the neutronic and thermal hydraulics aspects to propose a new core configuration to the IEA-R1 reactor of IPEN-CNEN/SP using U-Mo-Al fuels. Core configurations using U-10Mo-Al fuels with uranium densities variable from 3 to 8 gU/cm 3 were analyzed with the computational programs Citation and MTRCR-IEA R1. Core configurations for fuels with uranium densities variable from 3 to 5 gU/cm 3 showed to be adequate to use in IEA-R1 reactor e should present a stable in reactor performance even at high burn-up. (author)

  16. Study on thermo-oxide layers of uranium-niobium alloy

    International Nuclear Information System (INIS)

    Luo Lizhu; Yang Jiangrong; Zhou Ping

    2010-01-01

    Surface oxides structure of uranium-niobium alloys which were annealed under different temperatures (room temperature, 100, 200, 300 degree C, respectively)in air were studied by X-ray photoelectron spectroscopy (XPS) analysis and depth profile. Thickness of thermo-oxide layers enhance with the increasing oxide temperature, and obvious changes to oxides structure are observed. Under different delt temperatures, Nb 2 O 5 are detected on the initial surface of U-Nb alloys, and a layer of NbO mixed with some NbO x (0 2 O 5 and Nb metal. Dealing samples in air from room temperature to 200 degree C, non-stoichiometric UO 2+x (UO 2 + interstitial oxygen, P-type semiconductor) are found on initial surface of U-Nb alloys, which has 0.7 eV shift to lower binding energy of U 4f 7/2 characteristics comparing to that of UO 2 . Under room temperature, UO 2 are commonly detected in the oxides layer, while under temperature of 100 and 200 degree C, some P-type UO 2+x are found in the oxide layers,which has a satellite at binding energy of 396.6 eV. When annealing at 300 degree C, higher valence oxides, such as U 3 O 8 or UO x (2 5/2 and U 4f 7/2 peaks are 392.2 and 381.8 eV, respectively. UO 2 mixed uranium metal are the main compositions in the oxide layers. From the results, influence of temperature to oxidation of uranium is more visible than to niobium in uranium-niobium alloys. (authors)

  17. Vacuum-induction melting, refining, and casting of uranium and its alloys

    Energy Technology Data Exchange (ETDEWEB)

    Jackson, R J

    1989-10-11

    The vacuum-induction melting (VIM), refining, and casting of uranium and its alloys are discussed. Emphasis is placed on historical development, VIM equipment, crucible and mold design, furnace atmospheres, melting parameters, impurity pickup, ingot quality, and economics. The VIM procedures used to produce high-purity, high-quality sound ingots at the US Department of Energy Rocky Flats Plant are discussed in detail.

  18. An investigation of the γ → α martensitic transformation in uranium alloys

    International Nuclear Information System (INIS)

    Speer, J.G.; Edmonds, D.V.

    1988-01-01

    A detailed study of the γ → chi martensite transformation in uranium alloys is presented. Five binary uranium-base alloys containing 0.77 Ti, 1.2 Mo, 2.2 Mo, 4.3 Mo and 5.0 Mo, respectively, were examined. As quenched, the U-0.77 Ti and U-1.2 Mo alloys consisted of an orthorhombic α'/sub a/ martensite phase with an acicular morphology. The acicular martensite plates contain deformation twins which result from transformation stresses. The U-2.2 Mo and U-4.3 Mo alloys transformed during quenching to orthorhomic chi'/sub b/ and monoclinic chi'/sub b/ martensite phases, respectively. The banded morphology observed in these two alloys consists of long, parallel martensite plates containing fine arrays of transformation twins. The type I transformation twinning modes were identified as /021/, /130/ and /131/. There was also evidence for a type II /111/ mode. It was found that adjacent bands could contain different kinds of transformation twins. In the U-5.0 Mo alloy, some of the cubic parent phase was retained during water quenching, and chi/γ orientation relationship was determined. The γ phase was completely retained in this alloy by slow cooling from the solution treatment temperature of 800 0 C, and it was found that a martensitic reaction could be induced by deformation. The strain-induced martensite plates contained /021/ transformation twins. The chi/γ orientation relationship was found to be different than the one determined in the quenched condition, and both orientation relationships are irrational. The invariant plane strain theory of martensite crystallography was applied to the twinned martensites, and a number of different parent/product lattice correspondences were considered for the γ → chi transformations. It was concluded that more than one correspondence may be operative during these transformations

  19. Study of the quenching and subsequent return to room temperature of uranium-chromium, uranium-iron, and uranium-molybdenum alloys containing only small amounts of the alloying element; Etude de la trempe et du revenu a la temperature ordinaire d'alliages uranium-chrome, uranium-fer et uranium-molybdene, a faible teneur en element d'alliage

    Energy Technology Data Exchange (ETDEWEB)

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

    1960-09-15

    By means of an apparatus which makes possible thermal pre-treatments in vacuo, quenching carried out in a high purity argon atmosphere, and simultaneous recording of time temperature cooling and thermal contraction curves, the author has examined the transformations which occur in uranium-chromium, uranium-iron and uranium-molybdenum alloys during their quenching and subsequent return to room temperature. For uranium-chromium and uranium-iron alloys, the temperature at which the {gamma} {yields} {beta} transformation starts varies very little with the rate of cooling. For uranium-molybdenum alloys containing 2,8 atom per cent of Mo, this temperature is lowered by 120 deg. C for a cooling rate of 500 deg. C/mn. The temperature at which the {beta} {yields} {alpha} transformation starts is lowered by 170 deg. C for a cooling rate of 500 deg. C/mn in the case of uranium-chromium alloy containing 0,37 atom per cent of Cr. The temperature is little affected in the case of uranium-iron alloys. The addition of chromium or iron makes it possible to conserve the form {beta} at ordinary temperatures after quenching from the {beta} and {gamma} regions. The {beta} phase is particularly unstable and changes into needles of the {alpha} form even at room temperatures according to an autocatalytic transformation law similar to the austenitic-martensitic transformation law in the case of iron. The {beta} phase obtained by quenching from the {beta} phase region is more stable than that obtained by quenching from the {gamma} region. Chromium is a more effective stabiliser of the {beta} phase than is iron. Unfortunately it causes serious surface cracking. The {beta} {yields} {alpha} transformation in uranium-chromium alloys has been followed at room temperature by means of micro-cinematography. The author has not observed the direct {gamma} {yields} {alpha} transformation in uranium-molybdenum alloys containing 2,8 per cent of molybdenum even for cooling rates of up to 2000 deg. C

  20. Study of the quenching and subsequent return to room temperature of uranium-chromium, uranium-iron, and uranium-molybdenum alloys containing only small amounts of the alloying element; Etude de la trempe et du revenu a la temperature ordinaire d'alliages uranium-chrome, uranium-fer et uranium-molybdene, a faible teneur en element d'alliage

    Energy Technology Data Exchange (ETDEWEB)

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

    1960-09-15

    By means of an apparatus which makes possible thermal pre-treatments in vacuo, quenching carried out in a high purity argon atmosphere, and simultaneous recording of time temperature cooling and thermal contraction curves, the author has examined the transformations which occur in uranium-chromium, uranium-iron and uranium-molybdenum alloys during their quenching and subsequent return to room temperature. For uranium-chromium and uranium-iron alloys, the temperature at which the {gamma} {yields} {beta} transformation starts varies very little with the rate of cooling. For uranium-molybdenum alloys containing 2,8 atom per cent of Mo, this temperature is lowered by 120 deg. C for a cooling rate of 500 deg. C/mn. The temperature at which the {beta} {yields} {alpha} transformation starts is lowered by 170 deg. C for a cooling rate of 500 deg. C/mn in the case of uranium-chromium alloy containing 0,37 atom per cent of Cr. The temperature is little affected in the case of uranium-iron alloys. The addition of chromium or iron makes it possible to conserve the form {beta} at ordinary temperatures after quenching from the {beta} and {gamma} regions. The {beta} phase is particularly unstable and changes into needles of the {alpha} form even at room temperatures according to an autocatalytic transformation law similar to the austenitic-martensitic transformation law in the case of iron. The {beta} phase obtained by quenching from the {beta} phase region is more stable than that obtained by quenching from the {gamma} region. Chromium is a more effective stabiliser of the {beta} phase than is iron. Unfortunately it causes serious surface cracking. The {beta} {yields} {alpha} transformation in uranium-chromium alloys has been followed at room temperature by means of micro-cinematography. The author has not observed the direct {gamma} {yields} {alpha} transformation in uranium-molybdenum alloys containing 2,8 per cent of molybdenum even for cooling rates of up to 2000 deg. C

  1. Mechanical properties of depleted uranium-2 w/o molybdenum alloy

    International Nuclear Information System (INIS)

    Deel, O.L.; Burian, R.J.

    1979-01-01

    The primary objective of this program is to develop data and techniques for determining the dynamic impact response of radioactive-material shipping-container systems for environmental control and safety overview and assessment. One phase of this program is the dynamic testing of 1/8-, 1/4-, and 1/2-scale models of uranium-shielded truck casks. These linearly scaled models are fabricated from the same materials typically used in full-size prototype casks. In order to analytically evaluate the results of dynamic tests, it is necessary to know the mechanical properties of the materials of construction. Since the properties of cast uranium--molybdenum alloys vary significantly with casting and heat-treating techniques, it is necessary to fully characterize the mechanical properties of the uranium used in the model tests. This report presents the results of these studies. The uranium alloy exhibited a tensile strength equal to or greater than that reported by others. As indicated by the percentage of elongation and reduction in area, the ductility was lower. Comparative data for the other mechanical properties measured were not found in the literature

  2. Uranium determination in U-Al alloy with statistical tools support

    International Nuclear Information System (INIS)

    Furusawa, Helio Akira; Medalla, Felipe Quirino; Cotrim, Marycel Elena Barbosa; Pires, Maria Aparecida Faustino

    2011-01-01

    ICP-OES was used to quantify total uranium in natural UAl x powder alloy. A simple solubilisation procedure using diluted HNO 3 /HCl was successfully applied. Only 100 mg of sample were used which is an advantage over the volumetric methodologies. Only two dilutions were needed to reach measurable concentration. No other treatment was applied to the solutions. Calibration curves of three uranium lines (367.007, 385.958 and 409.014 nm) were evaluated using ANOVA. Comparing the indicators, the 367.007 nm line was the poorer one but exhibiting a R 2 = 0.998 and 0.9996 and 0.999 for the other two lines. No significant difference was found between these two lines. If needed, the 385.958 nm line could be used to quantify uranium in very low concentrations but with few advantages over the 409.014 nm line, if so. The average uranium concentration found was 0.80±0.01 μg.g-1, as expected for a predominant UAl 2 phase alloy. Higher uranium concentrations are also expected to be successfully quantified using these lines. In order to verify possibly inhomogeneity due to the high uranium concentration, one-way ANOVA was applied to 3 replicates. Homogeneity was confirmed measuring in both 385.958 and 409.014 nm lines. The uncertainty of solution homogeneity was estimated also in these two emission lines giving 0.006 and 0.005 μg.g-1, respectively. These two values are in compliance with the standard deviation of the average. (author)

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

    International Nuclear Information System (INIS)

    McDeavitt, Sean M.

    2011-01-01

    beginning of the materials processing setup. Also included within this section is a thesis proposal by Jeff Hausaman. Appendix C contains the public papers and presentations introduced at the 2010 American Nuclear Society Winter Meeting. Appendix A - MSNE theses of David Garnetti and Grant Helmreich and proposal by Jeff Hausaman A.1 December 2009 Thesis by David Garnetti entitled 'Uranium Powder Production Via Hydride Formation and Alpha Phase Sintering of Uranium and Uranium-Zirconium Alloys for Advanced Nuclear Fuel Applications' A.2 September 2009 Presentation by David Garnetti (same title as document in Appendix B.1) A.3 December 2010 Thesis by Grant Helmreich entitled 'Characterization of Alpha-Phase Sintering of Uranium and Uranium-Zirconium Alloys for Advanced Nuclear Fuel Applications' A.4 October 2010 Presentation by Grant Helmreich (same title as document in Appendix B.3) A.5 Thesis Proposal by Jeffrey Hausaman entitled 'Hot Extrusion of Alpha Phase Uranium-Zirconium Alloys for TRU Burning Fast Reactors' Appendix B - External presentations introduced at the 2010 ANS Winter Meeting B.1 J.S. Hausaman, D.J. Garnetti, and S.M. McDeavitt, 'Powder Metallurgy of Alpha Phase Uranium Alloys for TRU Burning Fast Reactors,' Proceedings of 2010 ANS Winter Meeting, Las Vegas, Nevada, USA, November 7-10, 2010 B.2 PowerPoint Presentation Slides from C.1 B.3 G.W. Helmreich, W.J. Sames, D.J. Garnetti, and S.M. McDeavitt, 'Uranium Powder Production Using a Hydride-Dehydride Process,' Proceedings of 2010 ANS Winter Meeting, Las Vegas, Nevada, USA, November 7-10, 2010 B.4. PowerPoint Presentation Slides from C.3 B.5 Poster Presentation from C.3 Appendix C - Fuel cycle research and development undergraduate materials and poster presentation C.1 Poster entitled 'Characterization of Alpha-Phase Sintering of Uranium and Uranium-Zirconium Alloys' presented at the Fuel Cycle Technologies Program Annual Meeting C.2 April 2011 Honors Undergraduate Thesis by William Sames, Research Fellow

  4. Solubility of uranium in liquid gallium, indium and their alloys

    International Nuclear Information System (INIS)

    Volkovich, Vladimir A.; Maltsev, Dmitry S.; Yamschikov, Leonid F.; Osipenko, Alexander G.; Kormilitsyn, Mikhail V.

    2014-01-01

    Pyrochemical reprocessing of spent nuclear fuels (SNF) employing molten salts and liquid metals as working media is considered as a possible alternative to the existing liquid extraction (PUREX) processes. Liquid salts and metals allow reprocessing highly irradiated high burn-up fuels with short cooling times, including the fuels of fast neutron reactors. Pyrochemical technology opens a way to practical realization of short closed fuel cycle. Liquid low-melting metals are immiscible with molten salts and can be effectively used for separation (or selective extraction) of SNF components dissolved in fused salts. Binary or ternary alloys of eutectic compositions can be employed to lower the melting point of the metallic phase. However, the information on SNF components behaviour and properties in ternary liquid metal alloys is very scarce

  5. Study on hydrogen absorption/desorption properties of uranium alloys

    Energy Technology Data Exchange (ETDEWEB)

    Ito, Hiroshi; Yamaguchi, Kenji; Yamawaki, Michio [Tokyo Univ., Tokai, Ibaraki (Japan). Nuclear Engineering Research Lab.

    1996-10-01

    Hydrogen absorption/desorption properties of two U-Mn intermetallic compounds, U{sub 6}Mn and UMn{sub 2}, were investigated. U{sub 6}Mn absorbed hydrogen and the hydrogen desorption pressure of U{sub 6}Mn obtained from this experiment was higher than that of U, which was considered to be the effect of alloying, whereas UMn{sub 2} was not observed to absorb hydrogen up to 50 atm at room temperature. (author)

  6. Processing and Applications of Depleted Uranium Alloy Products

    Science.gov (United States)

    1976-09-01

    ammunition, weapons, gyrorotors, and ballast. Depleted uranium used in fly- wheel devices, nuclear fuel casks, and ammunition could consume a significant...from straight in the range of 0,002 to 0.060-inch TIR (total indicated runout ) with an average of 0.025-inch TIR.* Solution heat treatment of the as-cast...an envelope thickness of 0.050 inch to allow for runout and to clean up surface imperfections. The runout resulting from heat treatment was in the

  7. Uranium-molybdenum alloys containing 0,5 to 3 per cent by weight of molybdenum

    International Nuclear Information System (INIS)

    Lehmann, J.

    1959-01-01

    The following properties have been determined in the new cast state of uranium alloys containing 0.5-1-1.8-2 and 3.5 per cent of molybdenum: micro-graphical aspect, crystalline structure, thermal expansion, the mechanical characteristics, behaviour when subjected to cyclic temperature variations, and heat treatment. The transformation curves have been established for continuous cooling at rates varying between 2.5 and 200 deg. C per minute, using a dilatation method for the alloys containing 1.0, 2.0 and 3.0 per cent Mo. T.T.T. curves have been traced for 0.5 and 1.0 per cent Mo alloys and the Ms points determined for alloys containing 2.0 and 3.0 par cent Mo. In this way it has been possible to show the different results of transformation, brought about either by nucleation and diffusion or by shear - the alloy containing 1 per cent Mo, give two martensites α' and α'' and the alloys containing 2 and 3 per cent Mo give one martensite with a band structure. (author) [fr

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

    Energy Technology Data Exchange (ETDEWEB)

    McDeavitt, Sean M

    2011-04-29

    outlining the beginning of the materials processing setup. Also included within this section is a thesis proposal by Jeff Hausaman. Appendix C contains the public papers and presentations introduced at the 2010 American Nuclear Society Winter Meeting. Appendix A—MSNE theses of David Garnetti and Grant Helmreich and proposal by Jeff Hausaman A.1 December 2009 Thesis by David Garnetti entitled “Uranium Powder Production Via Hydride Formation and Alpha Phase Sintering of Uranium and Uranium-Zirconium Alloys for Advanced Nuclear Fuel Applications” A.2 September 2009 Presentation by David Garnetti (same title as document in Appendix B.1) A.3 December 2010 Thesis by Grant Helmreich entitled “Characterization of Alpha-Phase Sintering of Uranium and Uranium-Zirconium Alloys for Advanced Nuclear Fuel Applications” A.4 October 2010 Presentation by Grant Helmreich (same title as document in Appendix B.3) A.5 Thesis Proposal by Jeffrey Hausaman entitled “Hot Extrusion of Alpha Phase Uranium-Zirconium Alloys for TRU Burning Fast Reactors” Appendix B—External presentations introduced at the 2010 ANS Winter Meeting B.1 J.S. Hausaman, D.J. Garnetti, and S.M. McDeavitt, “Powder Metallurgy of Alpha Phase Uranium Alloys for TRU Burning Fast Reactors,” Proceedings of 2010 ANS Winter Meeting, Las Vegas, Nevada, USA, November 7-10, 2010 B.2 PowerPoint Presentation Slides from C.1 B.3 G.W. Helmreich, W.J. Sames, D.J. Garnetti, and S.M. McDeavitt, “Uranium Powder Production Using a Hydride-Dehydride Process,” Proceedings of 2010 ANS Winter Meeting, Las Vegas, Nevada, USA, November 7-10, 2010 B.4. PowerPoint Presentation Slides from C.3 B.5 Poster Presentation from C.3 Appendix C—Fuel cycle research and development undergraduate materials and poster presentation C.1 Poster entitled “Characterization of Alpha-Phase Sintering of Uranium and Uranium-Zirconium Alloys” presented at the Fuel Cycle Technologies Program Annual Meeting C.2 April 2011 Honors Undergraduate Thesis

  9. Incentives for the use of depleted uranium alloys as transport cask containment structure

    International Nuclear Information System (INIS)

    McConnell, P.; Salzbrenner, R.; Wellman, G.W.; Sorenson, K.B.

    1992-01-01

    Radioactive material transport casks use either lead or depleted uranium (DU) as gamma-ray shielding material. Stainless steel is conventionally used for structural containment. If a DU alloy had sufficient properties to guarantee resistance to failure during both nominal use and accident conditions to serve the dual-role of shielding and containment, the use of other structure materials (i.e., stainless steel) could be reduced. (It is recognized that lead can play no structural role.) Significant reductions in cask weight and dimensions could then be achieved perhaps allowing an increase in payload. The mechanical response of depleted uranium has previously not been included in calculations intended to show that DU-shielded transport casks will maintain their containment function during all conditions. This paper describesa two-part study of depleted uranium alloys: First, the mechanical behavior of DU alloys was determined in order to extend the limited set of mechanical properties reported in the literature. The mechanical properties measured include the tensile behavior the impact energy. Fracture toughness testing was also performed to determine the sensitivity of DU alloys to brittle fracture. Fracture toughness is the inherent material property which quantifies the fracmm resistance of a material. Tensile strength and ductility are significant in terms of other failure modes, however, as win be discussed. These mechanical properties were then input into finite element calculations of cask response to loading conditions to quantify the potential for claiming structural credit for DU. (The term ''structural credit'' describes whether a material has adequate properties to allow it to assume a positive role in withstanding structural loadings.)

  10. Incentives for the use of depleted uranium alloys as transport cask containment structure

    International Nuclear Information System (INIS)

    McConnell, P.; Salzbrenner, R.; Wellman, G.W.; Sorenson, K.B.

    1993-01-01

    Radioactive material transport casks use either lead or depleted uranium (DU) as gamma-ray shielding material. Stainless steel is conventionally used for structural containment. If a DU alloy had sufficient properties to guarantee resistance to failure during both normal use and accident conditions to serve the dual-role of shielding and containment, the use of other structural materials (i.e., stainless steel) could be reduced. (It is recognized that lead can play no structural role.) Significant reductions in cask weight and dimensions could then be achieved perhaps allowing an increase in payload. The mechanical response of depleted uranium has previously not been included in calculations intended to show that DU-shielded transport casks will maintain their containment function during all conditions. This paper describes a two-part study of depleted uranium alloys: First, the mechanical behavior of DU alloys was determined in order to extend the limited set of mechanical properties reported in the literature (Eckelmeyer, 1991). The mechanical properties measured include the tensile behavior the impact energy. Fracture toughness testing was also performed to determine the sensitivity of DU alloys to brittle fracture. Fracture toughness is the inherent material property which quantifies the fracture resistance of a material. Tensile strength and ductility are significant in terms of other failure modes, however, as will be discussed. These mechanical properties were then input into finite element calculations of cask response to loading conditions to quantify the potential for claiming structural credit for DU. (The term 'structural credit' describes whether a material has adequate properties to allow it to assume a positive role in withstanding structural loadings.) (J.P.N.)

  11. The use of slightly alloyed uranium as fuel: its influence on the dissolution and other stages of treatment

    International Nuclear Information System (INIS)

    Faugeras, P.; Leroy, P.; Lheureux, C.

    1959-01-01

    This report deals chiefly with the treatment of binary alloys (UAI, UMo, UZr, UCr, USi) with a low concentration of the additional element (≤2 per cent). The investigation was pursued with a view to the continued utilisation, with a minimum of modification, of the existing plants for treatment of non-alloyed irradiated uranium. In the first part, the usual process for the treatment of irradiated uranium by solvent extraction is briefly recalled. The second part is devoted to a study of the selective dissolution of the canning around certain of these alloys. The third part gives the behaviour of these different alloys at various phases of the usual treatment: a) dissolution; b) extractions; c) final treatment of fission products; d) final purification of plutonium. To conclude, possible alloys are classed as a function of their repercussions on the normal treatment. (author) [fr

  12. Uranium

    International Nuclear Information System (INIS)

    Hamdoun, N.A.

    2007-01-01

    The article includes a historical preface about uranium, discovery of portability of sequential fission of uranium, uranium existence, basic raw materials, secondary raw materials, uranium's physical and chemical properties, uranium extraction, nuclear fuel cycle, logistics and estimation of the amount of uranium reserves, producing countries of concentrated uranium oxides and percentage of the world's total production, civilian and military uses of uranium. The use of depleted uranium in the Gulf War, the Balkans and Iraq has caused political and environmental effects which are complex, raising problems and questions about the effects that nuclear compounds left on human health and environment.

  13. Optimisation by plastic deformation of structural and mechanical uranium alloys properties

    International Nuclear Information System (INIS)

    Prunier, Claude.

    1981-08-01

    Structural and mechanical properties evolution of rich and poor uranium alloys are investigated. Good usual properties are obtained with few metallic additions with a limited effect giving a fine and isotrope grain structure. Amelioration is observed with heat treatment from β and γ phases high temperature range. However, dynamic recrystallisation, related to hot working, is the better phenomena to maximize the usual mechanical and structural properties. So high temperature behaviour of rich and poor uranium alloys in α, β and γ crystalline structure is studied: - dynamic recrystallisation phenomena begins only in α, and β phases high temperature range; - high strength and brittle β phase shows a very large ductility above 700 deg C. Recrystallisation is a thermal actived phenomena localised at grain boundary, dependant with alloys concentration and crystalline structure. β phase activation energy and deformation rate for dynamic recrystallisation beginning are most important, than α and γ phases in relation with quadratic structure complexity. Both temperature and deformation rate are the main dynamic recrystallisation factors. Optimal usual mechanical and structural properties obtained by hot working (forging, milling) are sensible to hydrogen embrittlement [fr

  14. Fermi energy 5f spectral weight variation in uranium alloys

    Energy Technology Data Exchange (ETDEWEB)

    Denlinger, J.D.; Clack, J.; Allen, J.W. [Univ. of Michigan, Ann Arbor, MI (United States)] [and others

    1997-04-01

    Uranium materials display a wide range of thermal, electrical and magnetic properties, often exotic. For more than a decade there have been efforts to use photoemission spectroscopy to develop a systematic and unified understanding of the 5f electron states giving rise to this behavior. These efforts have been hampered by a paucity of systems where changes in transport properties are accompanied by substantial spectral changes, so as to allow an attempt to correlate the two kinds of properties within some model. The authors have made resonant photoemission measurements to extract the 5f spectral weight in three systems which show varying degrees of promise of permitting such an attempt, Y{sub 1{minus}x}U{sub x}Pd{sub 3}, U(Pd{sub x}Pt{sub 1{minus}x}){sub 3} and U(Pd{sub x}Cu{sub 1{minus}x}){sub 5}. They have also measured U 4f core level spectra. The 4f spectra can be modeled with some success by the impurity Anderson model (IAM), and the 5f spectra are currently being analyzed in that framework. The IAM characterizes the 5f-electrons of a single site by an f binding energy {epsilon}{sub f}, an f Coulomb interaction and a hybridization V to conduction electrons. Latent in the model are the phenomena of 5f mixed valence and the Kondo effect.

  15. Thermal cycling behaviour and thermal stability of uranium-molybdenum alloys of low molybdenum content

    International Nuclear Information System (INIS)

    Decours, J.; Fabrique, B.; Peault, O.

    1963-01-01

    We have studied the behaviour during thermal cycling of as-cast U-Mo alloys whose molybdenum content varies from 0.5 to 3 per cent; results are given concerning grain stability during extended heat treatments and the effect of treatments combining protracted heating with thermal cycling. The thermal cycling treatments were carried out at 550, 575, 600 and 625 deg C for 1000 cycles; the protracted heating experiments were done at 550, 575, 600 and 625 deg C for 2000 hours (4000 hrs at 625 deg C). The 0.5 per cent alloy resists much better to the thermal cycling than does the non-alloyed uranium. This resistance is, however, much lower than that of alloys containing over l per cent, even at 550 deg C it improves after a heat treatment for grain-refining. Alloys of over 1.1 per cent have a very good resistance to a cycling treatment even at 625 deg C, and this behaviour improves with increasing concentrations up to 3 per cent. An increase in the temperature up to the γ-phase has few disadvantages provided that it is followed by rapid cooling (50 to 100 deg C/min). The α grain is fine, the γ-phase is of the modular form, and the behaviour during a thermal cycling treatment is satisfactory. If this cooling is slow (15 deg /hr) the α-grain is coarse and cycling treatment behaviour is identical to that of the 0.5 per cent alloy. The protracted heat treatments showed that the α-grain exhibits satisfactory stability after 2000 hours at 575, 600 and 625 deg C, and after 4000 hours at 625 deg C. A heat cycling treatment carried out after these tests affects only very little the behaviour of these alloys during cycling. (authors) [fr

  16. Irradiation testing of high density uranium alloy dispersion fuels

    International Nuclear Information System (INIS)

    Hayes, S.L.; Trybus, C.L.; Meyer, M.K.

    1997-10-01

    Two irradiation test vehicles have been designed, fabricated, and inserted into the Advanced Test Reactor in Idaho. Irradiation of these experiments began in August 1997. These irradiation tests were designed to obtain irradiation performance information on a variety of potential new, high-density dispersion fuels. Each of the two irradiation vehicles contains 32 microplates. Each microplate is aluminum clad, having an aluminum matrix phase and containing one of the following compositions as the fuel phase: U-10Mo, U-8Mo, U-6Mo, U-4Mo, U-9Nb-3Zr, U-6Nb-4Zr, U-5Nb-3Zr, U-6Mo-1Pt, U-6Mo-0.6Ru, U-10Mo-0.05Sn, U 2 Mo, or U 3 Si 2 . These experiments will be discharged at peak fuel burnups of 40% and 80%. Of particular interest is the fission gas retention/swelling characteristics of these new fuel alloys. This paper presents the design of the irradiation vehicles and the irradiation conditions

  17. Irradiation testing of high-density uranium alloy dispersion fuels

    International Nuclear Information System (INIS)

    Hayes, S.L.; Trybus, C.L.; Meyer, M.K.

    1997-01-01

    Two irradiation test vehicles have been designed, fabricated, and inserted into the Advanced Test Reactor in Idaho. Irradiation of these experiments began in August 1997. These irradiation tests were designed to obtain irradiation performance information on a variety of potential new, high-density dispersion fuels. Each of the two irradiation vehicles contains 32 'microplates'. Each microplate is aluminum clad, having an aluminum matrix phase and containing one of the following compositions as the fuel phase: U-10Mo, U-8Mo, U-6Mo, U-4Mo, U-9Nb-3Zr, U-6Nb-4Zr, U-5Nb-3Zr, U-6Mo-1Pt, U-6Mo-0.6Ru, U10Mo-0.05Sn, U2Mo, or U 3 Si 2 . These experiments will be discharged at peak fuel burnups of approximately 40 and 80 at.% U 235 . Of particular interest are the extent of reaction of the fuel and matrix phases and the fission gas retention/swelling characteristics of these new fuel alloys. This paper presents the design of the irradiation vehicles and the irradiation conditions. (author)

  18. Spectrographic analysis of uranium-molybdenum alloys; Analisis espectrografico de aleaciones uranio-molibdeno

    Energy Technology Data Exchange (ETDEWEB)

    Roca, M

    1967-07-01

    A spectrographic method of analysis has been developed for uranium-molybdenum alloys containing up to 10 % Mo. The carrier distillation technique, with gallium oxide and graphite as carriers, is used for the semiquantitative determination of Al, Cr, Fe, Ni and Si, involving the conversion of the samples into oxides. As a consequence of the study of the influence of the molybdenum on the line intensities, it is useful to prepare only one set of standards with 0,6 % MoO{sub 3}. Total burning excitation is used for calcium, employing two sets of standards with 0,6 and 7.5 MoO{sub 3}. (Author) 5 refs.

  19. 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

  20. Elastic-plastic waves in UV 0.2 Uranium alloy

    International Nuclear Information System (INIS)

    Bernier, H.; Lalle, P.

    1984-09-01

    Release waves coming from the back face of an uranium alloy projectile in a symmetric collision are used to estimate some dynamic characteristics of this material. In the pressure range experimentally covered (<=29GPa) the velocity of the elastic precursor is about 3,45 km/s, and the Hugoniot elastic limit (HEL) is 1,15GPa. The pressure decrease behind the 20GPa (29GPa) shock wave begins with a quasi-elastic wave which velocity is 3,9 km/s (4,2 km/s), and pressure jump of 3GPa (3,7GPa)

  1. Enbrittlement of the U-7,5 Nb-2,5 Zr uranium alloy in gaseous environments

    International Nuclear Information System (INIS)

    Lepoutre, D.

    1984-10-01

    Stress corrosion cracking in air, oxygen, hydrogen, water, carbon dioxide of an uranium alloy U 7.5 Nb 2.5 Zr is experimentally studied. The stress corrosion tests are performed with fatigue precracked Single Edge Notched specimens, and the Linear Elastic Fracture Mechanic concept is used to describe the stress state at the crack tip. The s.c.c. maps and the cracking kinetics are determined as a function of stress intensity factor, temperature and pressure. In oxygen, an embrittlement is observed in all the tests, for any temperature and pressure; cracking is transgranular and thermally activated. We propose a model which takes in account the concomitant buildup of an oxide film and niobium interfacial segregated zone. In hydrogen, an embrittlement is observed only at low pressure: hydriding occurs at high pressure. A brittle phase failure mechanism is proposed to explain the embrittling effect of hydrogen. Cracking in oxygen at low pressure is inhibited by water and carbon dioxide. Finally oxygen is the specie responsible for cracking in laboratory air [fr

  2. Uranium

    International Nuclear Information System (INIS)

    Cuney, M.; Pagel, M.; Leroy, J.

    1992-01-01

    First, this book presents the physico-chemical properties of Uranium and the consequences which can be deduced from the study of numerous geological process. The authors describe natural distribution of Uranium at different scales and on different supports, and main Uranium minerals. A great place in the book is assigned to description and classification of uranium deposits. The book gives also notions on prospection and exploitation of uranium deposits. Historical aspects of Uranium economical development (Uranium resources, production, supply and demand, operating costs) are given in the last chapter. 7 refs., 17 figs

  3. Monte Carlo criticality analysis of simple geometries containing tungsten-rhenium alloys engrained with uranium dioxide and uranium mononitride

    International Nuclear Information System (INIS)

    Webb, Jonathan A.; Charit, Indrajit

    2011-01-01

    Highlights: → The addition of rhenium to the tungsten matrix within W-UO 2 and W-UN CERMET materials can help reduce the risk of submersion criticality accidents while increasing the strength and ductility of tungsten based nuclear fuel elements. → The addition of rhenium up to 30 at.% to simple geometries containing W-UO 2 mixtures can increase the critical mass by 65 kg. → The addition of rhenium up to 30 at.% to simple geometries containing W-UN mixtures can increase the critical mass by 22 kg. → The addition of rhenium by up to 30 at.% to simple geometries containing W-UO 2 mixtures can reduce the change in reactivity change due to water submersion by $5.07. → The addition of rhenium by up to 30 at.% to simple geometries containing W-UN mixtures can reduce the change in reactivity due to water submersion by $3.24. - Abstract: The critical mass and dimensions of simple geometries containing highly enriched uranium dioxide (UO 2 ) and uranium mononitride (UN) encapsulated in tungsten-rhenium alloys are determined using MCNP5 criticality calculations. Spheres as well as cylinders with length to radius ratios of 1.82 are computationally built to consist of 60 vol.% fuel and 40 vol.% metal matrix. Within the geometries, the uranium is enriched to 93 wt.% uranium-235 and the rhenium content within the metal alloy was modeled over the range of 0-30 at.%. The spheres containing UO 2 were determined to have a critical radius of 18.29-19.11 cm and a critical mass ranging from 366 kg to 424 kg. The cylinders containing UO 2 were found to have a critical radius ranging from 17.07 cm to 17.84 cm with a corresponding critical mass of 406-471 kg. Spheres engrained with UN were determined to have a critical radius ranging from 14.82 cm to 15.19 cm and a critical mass between 222 kg and 242 kg. Cylinders which were engrained with UN were determined to have a critical radius ranging from 13.81 cm to 14.15 cm and a corresponding critical mass of 245-267 kg. The critical

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

    Energy Technology Data Exchange (ETDEWEB)

    Lagerberg, G

    1961-04-15

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

  5. A Study on the Fabrication of Uranium-Cadmium Alloy and its Distillation Behavior

    International Nuclear Information System (INIS)

    Kim, Ji Yong; Ahn, Do Hee; Kim, Kwang Rag; Paek, Seung Woo; Kim, Si Hyung

    2010-01-01

    The pyrometallurgical nuclear fuel recycle process, called pyroprocessing, has been known as a promising nuclear fuel recycling technology. Pyroprocessing technology is crucial to advanced nuclear systems due to increased nuclear proliferation resistance and economic efficiency. The basic concept of pyroprocessing is group actinide recovery, which enhances the nuclear proliferation resistance significantly. One of the key steps in pyroprocessing is 'electrowinning' which recovers group actinides with lanthanide from the spent nuclear fuels. In this study, a vertical cadmium distiller was manufactured. The evaporation rate of pure cadmium in vertical cadmium distiller varied from 12.3 to 40.8 g/cm 2 /h within a temperature range of 773 ∼ 923 K and pressure below 0.01 torr. Uranium - cadmium alloy was fabricated by electrolysis using liquid cadmium cathode in a high purity argon atmosphere glove box. The distillation behavior of pure cadmium and cadmium in uranium - cadmium alloy was investigated. The distillation behavior of cadmium from this study could be used to develop an actinide recovery process from a liquid cadmium cathode in a cadmium distiller

  6. A Study on the Fabrication of Uranium-Cadmium Alloy and its Distillation Behavior

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Ji Yong [University of Science and Technology, Daejeon (Korea, Republic of); Ahn, Do Hee; Kim, Kwang Rag; Paek, Seung Woo; Kim, Si Hyung [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2010-12-15

    The pyrometallurgical nuclear fuel recycle process, called pyroprocessing, has been known as a promising nuclear fuel recycling technology. Pyroprocessing technology is crucial to advanced nuclear systems due to increased nuclear proliferation resistance and economic efficiency. The basic concept of pyroprocessing is group actinide recovery, which enhances the nuclear proliferation resistance significantly. One of the key steps in pyroprocessing is 'electrowinning' which recovers group actinides with lanthanide from the spent nuclear fuels. In this study, a vertical cadmium distiller was manufactured. The evaporation rate of pure cadmium in vertical cadmium distiller varied from 12.3 to 40.8 g/cm{sup 2}/h within a temperature range of 773 {approx} 923 K and pressure below 0.01 torr. Uranium - cadmium alloy was fabricated by electrolysis using liquid cadmium cathode in a high purity argon atmosphere glove box. The distillation behavior of pure cadmium and cadmium in uranium - cadmium alloy was investigated. The distillation behavior of cadmium from this study could be used to develop an actinide recovery process from a liquid cadmium cathode in a cadmium distiller.

  7. 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.

  8. 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.

  9. Relationship Between Crack Growth Resistance KR Curve and Specimen Width for 2060 - T8E30 Lithium Aluminum Alloy

    Directory of Open Access Journals (Sweden)

    Tong Di Hua

    2016-01-01

    Full Text Available KR crack growth resistance curve can be used to predict crack propagation behavior, estimate the crack component bearing capacity after the crack, so KR curve research occupies very important position in the fracture mechanics. Based on crack growth resistance KR test curve of 2060 - T8E30 lithium aluminum alloy under the same thickness for different width, studies have shown that under the same thickness, the influence of the width on the resistance curve of crack propagation can be neglected. Empirical equation of resistance curve of crack extension of the smaller width specimen is given. Extending the fitting equation to that of larger width, it can be found that it is highly coincided with the experimental results.

  10. Properties of low content uranium-molybdenum alloys which may be used as nuclear fuels

    International Nuclear Information System (INIS)

    Lehmann, J.; Decours, J.

    1964-01-01

    Metallurgical properties are given in this report of uranium-molybdenum alloys containing 0,5 to 3 per cent of molybdenum. Since some of these alloys are used in EDF power reactors are given: briefly the operating conditions imposed on nuclear fuels: maximum temperature, temperature gradient and external pressure. In the first part are considered the structural properties of the alloys correlation with the phase transformation kinetics; a description is given of the effects of certain physico-metallurgical factors on the morphology and the crystalline structure of the materials: - solidification conditions and the heredity of the γ structure, - cooling rate at the transformation points, - whether or not the intermediate γ → β transformation is suppressed In the second part we show how a knowledge of the phase transformation processes has made it possible to define the optimum preparation conditions for these materials in the form of fuel tubes intended for the EDF reactors: casting conditions, controlled cooling treatments, weldability. In the third part we study the thermal, stability during the long duration high temperature treatments and the cycles in the two zones of the diagram α + γ; β + γ the effects of the morphology (in particular the two types of α pseudo-grains observed) and of the cooling rate during the transformation point transitions are described. In the fourth part are discussed the mechanical properties: resistance to a tractive force, resistance to creep, resilience. These properties can also be affected by the γ structure heredity and by the cooling rate to which the alloy has been subjected. In conclusion we discuss the reasons which led to the choice of some of these alloys for the first EDF reactors in particular the advantages of their high creep resistance between 450 and 600 deg C for use in the form of tubes subjected to an external pressure. (authors) [fr

  11. Melting, casting, and alpha-phase extrusion of the uranium-2.4 weight percent niobium alloy

    International Nuclear Information System (INIS)

    Anderson, R.C.; Beck, D.E.; Kollie, T.G.; Zorinsky, E.J.; Jones, J.M.

    1981-10-01

    The experimental details of the melting, casting, homogenization, and alpha-phase extrusion process used to fabricate the uranium-2.4 wt % niobium alloy into 46-mm-diameter rods is described. Extrusion defects that were detected by an ultrasonic technique were eliminated by proper choice of extrusion parameters; namely, reduction ratio, ram speed, die angle, and billet preheat temperature

  12. High-strength uranium-0.8 weight percent titanium alloy penetrators

    International Nuclear Information System (INIS)

    Northcutt, W.G.

    1978-09-01

    Long-rod kinetic-energy penetrators, produced from a uranium-0.8 titanium (U-0.8 Ti) alloy, are normally water quenched from the gamma phase (approximately 800 0 C) and aged to the desired hardness and strength levels. High cooling rates from 800 0 C in U-0.8 Ti alloy cylindrical bodies larger than about 13 mm in diameter cause internal voids, while slower rates of cooling can produce material that is unresponsive to aging. For the present study, elimination of quenching voids was of paramount importance; therefore, a process including the quenching of plate was explored. Vacuum-induction-cast ingots were forged and rolled into plate and cut into blanks from which the penetrators were obtained. Quenched U-0.8 Ti alloy blanks were aged at 350 to 500 0 C to determine the treatment that would provide maximum tensile and impact strengths. Both tensile and impact strengths were maximized by aging in vacuum for six hours at 450 0 C

  13. Uranium

    International Nuclear Information System (INIS)

    Anon.

    1984-01-01

    The article briefly discusses the Australian government policy and the attitude of political party factions towards the mining and exporting of the uranium resources in Australia. Australia has a third of the Western World's low-cost uranium resources

  14. Uranium

    International Nuclear Information System (INIS)

    Poty, B.; Cuney, M.; Bruneton, P.; Virlogeux, D.; Capus, G.

    2010-01-01

    With the worldwide revival of nuclear energy comes the question of uranium reserves. For more than 20 years, nuclear energy has been neglected and uranium prospecting has been practically abandoned. Therefore, present day production covers only 70% of needs and stocks are decreasing. Production is to double by 2030 which represents a huge industrial challenge. The FBR-type reactors technology, which allows to consume the whole uranium content of the fuel, is developing in several countries and will ensure the long-term development of nuclear fission. However, the implementation of these reactors (the generation 4) will be progressive during the second half of the 21. century. For this reason an active search for uranium ores will be necessary during the whole 21. century to ensure the fueling of light water reactors which are huge uranium consumers. This dossier covers all the aspects of natural uranium production: mineralogy, geochemistry, types of deposits, world distribution of deposits with a particular attention given to French deposits, the exploitation of which is abandoned today. Finally, exploitation, ore processing and the economical aspects are presented. Contents: 1 - the uranium element and its minerals: from uranium discovery to its industrial utilization, the main uranium minerals (minerals with tetravalent uranium, minerals with hexavalent uranium); 2 - uranium in the Earth's crust and its geochemical properties: distribution (in sedimentary rocks, in magmatic rocks, in metamorphic rocks, in soils and vegetation), geochemistry (uranium solubility and valence in magmas, uranium speciation in aqueous solution, solubility of the main uranium minerals in aqueous solution, uranium mobilization and precipitation); 3 - geology of the main types of uranium deposits: economical criteria for a deposit, structural diversity of deposits, classification, world distribution of deposits, distribution of deposits with time, superficial deposits, uranium

  15. Contribution to the micrographic study of uranium and its alloys; Contribution a l'etude micrographique de l'uranium et de ses alliages

    Energy Technology Data Exchange (ETDEWEB)

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

    1956-06-15

    The present report is the result of research carried out by the radio metallurgy section, to perfect micrographic techniques applicable to the study of samples of irradiated uranium. In the first part of this work, two polishing baths are developed, having the qualities with a minimum of disadvantages inherent in their respective compositions: they are, on the one hand perchloric acid-ethanol mixtures, and on the other hand a phospho-chromic-ethanol bath. In the chapter following, the micrographic attack of uranium is studied. The only satisfactory process is oxidation by cathode bombardment forming epitaxic layers. In the third chapter, an attempt is made to characterise the different surface states of the uranium by dissolution potential measurements and electronic diffraction. In the fourth chapter are given some examples of the application of these techniques to the micrographic study of various uranium alloys. In an appendix, it is shown how the chemical oxidation after phospho-chromic-alcohol polishing allows the different inclusions present in the molten uranium to be distinguished. By X-ray diffraction, uranium monocarbide and mononitride inclusions in particular are characterised. (author) [French] Le present rapport est le resultat de recherches effectuees au service de radiometallurgie pour la mise au point de techniques micrographiques applicables a l'etude d'echantillons d'uranium irradie. Dans la premiere partie de ce travail, nous mettons au point deux bains de polissage qui presentent les qualites inherentes a leur composition respective, avec le minimum d'inconvenients: ce sont d'une part des melanges acide perchlorique-ethanol, et d'autre part un bain phospho-chromique-ethanol. Dans le chapitre suivant, nous etudions l'attaque micrographique de l'uranium. Seul le procede d'oxydation par bombardement cathodique formant des couches epitaxiques, est satisfaisant. Dans le troisieme chapitre, nous essayons de caracteriser les differents etats de

  16. Contribution to the micrographic study of uranium and its alloys; Contribution a l'etude micrographique de l'uranium et de ses alliages

    Energy Technology Data Exchange (ETDEWEB)

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

    1956-06-15

    The present report is the result of research carried out by the radio metallurgy section, to perfect micrographic techniques applicable to the study of samples of irradiated uranium. In the first part of this work, two polishing baths are developed, having the qualities with a minimum of disadvantages inherent in their respective compositions: they are, on the one hand perchloric acid-ethanol mixtures, and on the other hand a phospho-chromic-ethanol bath. In the chapter following, the micrographic attack of uranium is studied. The only satisfactory process is oxidation by cathode bombardment forming epitaxic layers. In the third chapter, an attempt is made to characterise the different surface states of the uranium by dissolution potential measurements and electronic diffraction. In the fourth chapter are given some examples of the application of these techniques to the micrographic study of various uranium alloys. In an appendix, it is shown how the chemical oxidation after phospho-chromic-alcohol polishing allows the different inclusions present in the molten uranium to be distinguished. By X-ray diffraction, uranium monocarbide and mononitride inclusions in particular are characterised. (author) [French] Le present rapport est le resultat de recherches effectuees au service de radiometallurgie pour la mise au point de techniques micrographiques applicables a l'etude d'echantillons d'uranium irradie. Dans la premiere partie de ce travail, nous mettons au point deux bains de polissage qui presentent les qualites inherentes a leur composition respective, avec le minimum d'inconvenients: ce sont d'une part des melanges acide perchlorique-ethanol, et d'autre part un bain phospho-chromique-ethanol. Dans le chapitre suivant, nous etudions l'attaque micrographique de l'uranium. Seul le procede d'oxydation par bombardement cathodique formant des couches epitaxiques, est satisfaisant. Dans le troisieme chapitre, nous essayons

  17. Uranium

    International Nuclear Information System (INIS)

    Mackay, G.A.

    1978-01-01

    The author discusses the contribution made by various energy sources in the production of electricity. Estimates are made of the future nuclear contribution, the future demand for uranium and future sales of Australian uranium. Nuclear power growth in the United States, Japan and Western Europe is discussed. The present status of the six major Australian uranium deposits (Ranger, Jabiluka, Nabarlek, Koongarra, Yeelerrie and Beverley) is given. Australian legislation relevant to the uranium mining industry is also outlined

  18. Uranium

    International Nuclear Information System (INIS)

    1982-01-01

    The development, prospecting, research, processing and marketing of South Africa's uranium industry and the national policies surrounding this industry form the headlines of this work. The geology of South Africa's uranium occurences and their positions, the processes used in the extraction of South Africa's uranium and the utilisation of uranium for power production as represented by the Koeberg nuclear power station near Cape Town are included in this publication

  19. Uranium

    International Nuclear Information System (INIS)

    Stewart, E.D.J.

    1974-01-01

    A discussion is given of uranium as an energy source in The Australian economy. Figures and predictions are presented on the world supply-demand position and also figures are given on the added value that can be achieved by the processing of uranium. Conclusions are drawn about Australia's future policy with regard to uranium (R.L.)

  20. Uranium

    International Nuclear Information System (INIS)

    Toens, P.D.

    1981-03-01

    The geological setting of uranium resources in the world can be divided in two basic categories of resources and are defined as reasonably assured resources, estimated additional resources and speculative resources. Tables are given to illustrate these definitions. The increasing world production of uranium despite the cutback in the nuclear industry and the uranium requirements of the future concluded these lecture notes

  1. Analysis of degradation in nickel-based alloys using focused ion beam imaging and specimen preparation combined with analytical electron microscopy

    International Nuclear Information System (INIS)

    Phaneuf, M.W.; Botton, G.A.

    2002-01-01

    Focused ion beam (FIB) microscopes have become well-established in the semiconductor industry during the past decade, and are rapidly gaining attention in the field of materials science, both as a tool for producing site specific, parallel sided transmission electron microscope (TEM) specimens and as stand alone specimen preparation and imaging systems. FIB secondary electron imaging (SEI) of nickel-based alloys, such as commercially produced Alloy 600 (approximately Ni 15Cr 10Fe 0.5C), has been demonstrated to show a high degree of sensitivity to the presence of deformation in the alloy, and FIB secondary ion imaging (SII) is particularly useful for identifying the presence of grain boundary corrosion, as secondary ion yields from metallic specimens can increase by three orders of magnitude in the presence of oxygen. This 'oxygen enhanced yield', makes FIB SII ideal for detection of corrosion at grain boundaries down to thicknesses of only a few tens of nanometers. Historically, while TEM has been considered the tool of choice for high resolution chemical and crystallographic analysis of specimens, the technique has suffered from difficulties in producing suitable samples from site-specific areas with a high probability of success. The advent of FIB specimen preparation for TEM has largely changed that. FIB imaging can be combined with FIB 'nano-machining' techniques to produce site-specific, parallel sided TEM specimens well-suited to analytical electron microscopy (AEM) analyses in the TEM, including electron energy loss spectroscopy (EELS), energy dispersive x-ray spectroscopy (EDX) and electron diffraction. When combined with new FIB-based methodologies for surveying large areas to exactly select the regions of interest, such as crack tips or the maximum extent of penetration of intergranular attack (IGA), subsequent FIB TEM specimen preparation and TEM analysis unite to produce a powerful tool to study these phenomena. Examples of these applications of FIB

  2. Basic design of a rotating disk centrifugal atomizer for uranium-molybdenum alloys

    International Nuclear Information System (INIS)

    Alzari, Silvio

    2001-01-01

    One of the most used techniques to produce metallic powders is the centrifugal atomization with a rotating disk. This process is employ to fabricate ductile metallic particles of uranium-molybdenum alloys (typically U- 7 % Mo, by weight) for nuclear fuel elements for research and testing reactors. These alloys exhibit a face-centered cubic structure (γ phase) which is stable above 700 C degrees and can be retained at room temperature. The rotating disk centrifugal atomization allows a rapid solidification of spherical metallic droplets of about 40 to 100 μm, considered adequate to manufacture nuclear fuel elements. Besides the thermo-physical properties of both the alloy and the cooling gas, the main parameters of the process are the radius of the disk (R), the diameter of the atomization chamber (D), the disk rotation speed (ω), the liquid volume flow rate (Q) and the superheating of the liquid (ΔT). In this work, they were applied approximate analytical models to estimate the optimal geometrical and operative parameters to obtain spherical metallic powder of U- 7 % Mo alloy. Three physical phenomena were considerate: the liquid metal flow along the surface of the disk, the fragmentation and spheroidization of the droplets and the cooling and solidification of the droplets. The principal results are the more suitable gas is helium; R ≅ 20 mm; D ≥ 1 m; ≅ 20,000 - 50,000 rpm; Q ≅ 4 - 10 cm 3 /s; ΔT ≅ 100 - 200 C degrees. By applying the relevant non-dimensional parameters governing the main physical phenomena, the conclusion is that the more appropriate non-radioactive metal to simulate the atomization of U- 7 % Mo is gold [es

  3. Development of an environmentally friendly protective coating for the depleted uranium-0.75 wt% titanium alloy

    International Nuclear Information System (INIS)

    Roeper, Donald F.; Chidambaram, Devicharan; Clayton, Clive R.; Halada, Gary P.; Derek Demaree, J.

    2006-01-01

    Molybdenum oxide-based conversion coatings have been formed on the surface of the depleted uranium-0.75 wt% titanium alloy using either concentrated nitric acid or fluorides for surface activation prior to coating formation. The acid-activated surface forms a coating that offers corrosion protection after a period of aging, when uranium species have migrated to the surface. X-ray photoelectron spectroscopy (XPS) revealed that the protective coating is primarily a polymolybdate bound to a uranyl ion. Rutherford backscattering spectroscopy (RBS) on the acid-activated coatings also shows uranium dioxide migrating to the surface. The fluoride-activated surface does not form a protective coating and there are no uranium species on the surface as indicated by XPS. The coating on the fluoride-activated samples has been found to contain a mixture of molybdenum oxides of which the main component is molybdenum trioxide and a minor component of an Mo(V) oxide

  4. Influence of specimen size and grain orientation to the life of a polycrystalline Ni-base alloy at LCF stress

    International Nuclear Information System (INIS)

    Seibel, Thomas

    2014-01-01

    In the present work the LCF (Low Cycle Fatigue) crack initiation life of the conventionally cast Ni-base alloy RENE 80 was analyzed as a function of specimen size and grain orientation. Five specimen geometries with distinctly different gauge sections were used: 3 geometries with cylindrical gauge section (G1-G3) and two notched geometries with a stress concentration factor of α 1 = 1,62 (KG1) and α 2 = 2,60 (KG2), resulting in a maximum difference of the damage relevant surface area up to a factor of approximately 72. Correction factors were determined by FEM calculations for all specimen geometries with highly reduced gauge sections where direct strain measurement was not possible. Additionally a uniform failure criterion with a relatively small crack size of 0,962 mm 2 was defined. Totally, 116 isothermal LCF tests were carried out at the different specimen types at a temperature of 850 C in total strain control with a load ratio (minimum strain / maximum strain) of R ε = -1. The load cycles were applied with triangular waveform at a frequency of 0.1 Hz for high strain amplitudes and 1 Hz for low strain amplitudes, respectively. After the LCF-Tests the fracture surfaces of all samples were analyzed in more detail by SEM to identify the crack initiation mechanisms as well as the crack initiation sites. In this context it could be shown, that fatigue cracks were generally initiated at slip bands in surface grains. Accordingly, the grain orientations at the crack initiation sites were measured by electron back scatter diffraction (EBSD) and the maximum shear stresses in the respective principal slip system (111) <110> was calculated using the Schmid approach. For this, longitudinal sections were be prepared exactly at the crack initiation sites of samples loaded with low strain amplitudes where clearly defined single crack initiation sites were observed. Afterwards the maximum shear stress in the principal slip system at the crack initiation site was correlated

  5. β → α isothermal transformation in pure and weakly alloyed uranium

    International Nuclear Information System (INIS)

    Aubert, H.; Lelong, C.

    1966-01-01

    The TTT diagrams describing the β → α isothermal transformation have been made by isothermal dilatometry for pure uranium and 21 alloys based on chromium, silicon, molybdenum, iron, aluminium, zirconium. The thermal cycle preceding the isothermal step influences the decomposition kinetics at temperature corresponding to the eutectoid and martensitic mechanisms, but not in the range where the bainitic transformation occurs. The stability of the β phase decreases with the chromium, molybdenum and silicon concentration: it is affected differently for each of the three transformation mechanisms. The ternary additions, even at very low concentration have a considerable effect on the stability. When the concentration decreases the martensitic mechanism is active at progressively higher temperature, diminishing to the point of disappearance the temperature range where the transformation is considered as being of the bainitic mode. (author) [fr

  6. First-principles investigations of the physical properties of binary uranium silicide alloys

    International Nuclear Information System (INIS)

    Yang, Jin; Long, Jianping; Yang, Lijun; Li, Dongmei

    2013-01-01

    Graphical abstract: Total density of states for USi 2 . Display Omitted -- Abstract: The structural, elastic properties and the Debye temperature of binary Uranium Silicide (U-Si) alloys are investigated by using the first-principles plane-wave pseudopotential density function theory within the generalized gradient approximation (GGA). The ground states properties are found to agree with the available experimental data. The mechanical properties like shear modulus, Young’s modulus, Poisson’s ratio σ and ratio B/G are also calculated. Finally, The averaged sound velocity (v m ), the longitudinal sound velocity (v l ), transverse sound velocity (v t ) and the Debye temperature (θ D ) are obtained. However, the theoretical values are slightly different from few existed experiment data because the latter was obtained at room temperature while the former one at 0 K

  7. Nuclear reactor fuel structure containing uranium alloy wires embedded in a metallic matrix plate

    International Nuclear Information System (INIS)

    Travelli, A.

    1988-01-01

    A nuclear fuel-containing plate structure for a nuclear reactor is described; such structure comprising a pair of malleable metallic non-fissionable matrix plates having confronting surfaces which are pressure bonded together and fully united to form a bonded surface, and elongated malleable wire-like fissionable fuel members separately confined and fully enclosed between the matrix plates along the interface to afford a high fuel density as well as structural integrity and effective retention of fission products. The plates have separate recesses formed in the confronting surfaces for closely receiving the wire-like fissionable fuel members. The wire-like fissionable fuel members are made of a maleable uranium alloy capable of being formed into elongated wire-like members and capable of withstanding pressure bonding. The wire-like fissionable fuel members are completely separated and isolated by fully united portions of the interface

  8. Determination of ultratrace amounts of uranium and thorium in aluminium and aluminium alloys by electrothermal vaporization/ICP-MS

    International Nuclear Information System (INIS)

    Nakamura, Yasushi; Kobayashi, Yoshio; Kakurai, Yousuke

    1993-01-01

    A method has been developed for determining the 0.01 ng g -1 level of uranium and thorium in aluminium and aluminium alloys by electrothermal vaporization (ETV)/ICP-MS. This method was found to be significantly interfered with any matrices or other elements contained. An ion-exchange technique was therefore applied to separate uranium and thorium from aluminium and other elements. It was known that uranium are adsorbed on an anion-exchange resin and thorium are adsorbed on cation-exchange resin. However, aluminium and copper were eluted with 6 M hydrochloric acid. Dissolve the sample with hydrochloric acid containing copper which was added for analysis of pure aluminium, and oxidize with hydrogen peroxide. Concentration of hydrochloric acid in the solution was adjusted to 6 M, and then passed the solution through the mixed ion-exchange resin column. After the uranium and thorium were eluted with 1 M hydrofluoric acid-0.1 M hydrochloric acid, the solution was evaporated to dryness. It was then dissolved with 1 M hydrochloric acid. Uranium and thorium were analyzed by ETV/ICP-MS using tungsten and molybdenum boats, respectively, since the tungsten boat contained high-level thorium and the molybdenum boat contained uranium. The determination limit of uranium and thorium were 0.003 and 0.005 ng g -1 , respectively. (author)

  9. Oxidation behavior of U-2wt%Nb, Ti, and Ni alloys in air

    International Nuclear Information System (INIS)

    Ju, J. S.; Yoo, K. S.; Jo, I. J.; Gug, D. H.; Su, H. S.; Lee, E. P.; Bang, K. S.; Kim, H. D.

    2003-01-01

    For the long term storage safety study of the metallic spent fuel, U-Nb, U-Ti, U-Ni, U-Zr, and U-Hf simulated metallic uranium alloys, known as corrosion resistant alloys, were fabricated and oxidized in oxygen gas at 200 .deg. C-300 .deg. C. Simulated metallic uranium alloys were more corrosion resistant than pure uranium metal, and corrosion resistance increases Nb, Ni, Ti in that order. The oxidation rates of uranium alloys determined and activation energy was calculated for each alloy. The matrix microstructure of the test specimens were analyzed using OM, SEM, and EPMA. It was concluded that Nb was the best acceptable alloying elements for reducing corrosion of uranium metal considered to suitable as candidate

  10. 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

  11. Development of a high density fuel based on uranium-molybdenum alloys with high compatibility in high temperatures

    International Nuclear Information System (INIS)

    Oliveira, Fabio Branco Vaz de

    2008-01-01

    This work has as its objective the development of a high density and low enriched nuclear fuel based on the gamma-UMo alloys, for utilization where it is necessary satisfactory behavior in high temperatures, considering its utilization as dispersion. For its accomplishment, it was started from the analysis of the RERTR ('Reduced Enrichment for Research and Test Reactors') results and some theoretical works involving the fabrication of gamma-uranium metastable alloys. A ternary addition is proposed, supported by the properties of binary and ternary uranium alloys studied, having the objectives of the gamma stability enhancement and an ease to its powder fabrication. Alloys of uranium-molybdenum were prepared with 5 to 10% Mo addition, and 1 and 3% of ternary, over a gamma U7Mo binary base alloy. In all the steps of its preparation, the alloys were characterized with the traditional techniques, to the determination of its mechanical and structural properties. To provide a process for the alloys powder obtention, its behavior under hydrogen atmosphere were studied, in thermo analyser-thermo gravimeter equipment. Temperatures varied from the ambient up to 1000 deg C, and times from 15 minutes to 16 hours. The results validation were made in a semi-pilot scale, where 10 to 50 g of powders of some of the alloys studied were prepared, under static hydrogen atmosphere. Compatibility studies were conducted by the exposure of the alloys under oxygen and aluminum, to the verification of possible reactions by means of differential thermal analysis. The alloys were exposed to a constant heat up to 1000 deg C, and their performances were evaluated in terms of their reaction resistance. On the basis of the results, it was observed that ternary additions increases the temperatures of the reaction with aluminum and oxidation, in comparison with the gamma UMo binaries. A set of conditions to the hydration of the alloys were defined, more restrictive in terms of temperature, time and

  12. Analysis of surface roughening behavior of 6063 aluminum alloy by tensile testing of a trapezoidal uniaxial specimen

    Energy Technology Data Exchange (ETDEWEB)

    Cai, Yang [School of Materials Science and Engineering, Harbin Institute of Technology, Harbin 150090 (China); Wang, Xiaosong, E-mail: hitxswang@hit.edu.cn [School of Materials Science and Engineering, Harbin Institute of Technology, Harbin 150090 (China); National Key Laboratory of Precision Hot Processing of Metals, Harbin Institute of Technology, Harbin 150001 (China); Yuan, Shijian [School of Materials Science and Engineering, Harbin Institute of Technology, Harbin 150090 (China); National Key Laboratory of Precision Hot Processing of Metals, Harbin Institute of Technology, Harbin 150001 (China)

    2016-08-30

    To determine the quantitative relationship between surface roughness and strain, the surface roughening behavior of a 6063 aluminum alloy tube was examined by tensile testing of a trapezoidal uniaxial specimen, that can provide a continuous strain distribution after tensile deformation. The surface roughness was measured using a laser scanning confocal microscope to reflect the degree of roughening. The microstructure and surface morphology were examined using electron back-scattered diffraction and in-situ scanning electron microscopy to determine the grain orientation and surface topography evolution. The surface roughness increased with strain when the strain was less than 0.067 and then decreased slightly, with a maximum surface roughness of 23.73 µm. Inhomogeneous deformation at the grain boundaries and inside the grains was enhanced with increasing strain, resulting in an increase of surface roughness when the strain was below a critical value. As the strain increased, a greater number of slip systems contributed to the further deformation. Thus, the strain became more homogeneous, and accordingly, the surface roughness slightly decreased.

  13. Model of the Evolution of Deformation Defects and Irreversible Strain at Thermal Cycling of Stressed TiNi Alloy Specimen

    Directory of Open Access Journals (Sweden)

    Volkov Aleksandr E.

    2015-01-01

    Full Text Available This microstructural model deals with simulation both of the reversible and irreversible deformation of a shape memory alloy (SMA. The martensitic transformation and the irreversible deformation due to the plastic accommodation of martensite are considered on the microscopic level. The irreversible deformation is described from the standpoint of the plastic flow theory. Isotropic hardening and kinematic hardening are taken into account and are related to the densities of scattered and oriented deformation defects. It is supposed that the phase transformation and the micro plastic deformation are caused by the generalized thermodynamic forces, which are the derivatives of the Gibbs’ potential of the two-phase body. In terms of these forces conditions for the phase transformation and for the micro plastic deformation on the micro level are formulated. The macro deformation of the representative volume of the polycrystal is calculated by averaging of the micro strains related to the evolution of the martensite Bain’s variants in each grain comprising this volume. The proposed model allowed simulating the evolution of the reversible and of the irreversible strains of a stressed SMA specimen under thermal cycles. The results show a good qualitative agreement with available experimental data. Specifically, it is shown that the model can describe a rather big irreversible strain in the first thermocycle and its fast decrease with the number of cycles.

  14. Influence of heat input on weld bead geometry using duplex stainless steel wire electrode on low alloy steel specimens

    Directory of Open Access Journals (Sweden)

    Ajit Mondal

    2016-12-01

    Full Text Available Gas metal arc welding cladding becomes a popular surfacing technique in many modern industries as it enhances effectively corrosion resistance property and wear resistance property of structural members. Quality of weld cladding may be enhanced by controlling process parameters. If bead formation is found acceptable, cladding is also expected to be good. Weld bead characteristics are often assessed by bead geometry, and it is mainly influenced by heat input. In this paper, duplex stainless steel E2209 T01 is deposited on E250 low alloy steel specimens with 100% CO2 gas as shielding medium with different heats. Weld bead width, height of reinforcement and depth of penetration are measured. Regression analysis is done on the basis of experimental data. Results reveal that within the range of bead-on-plate welding experiments done, parameters of welding geometry are on the whole linearly related with heat input. A condition corresponding to 0.744 kJ/mm heat input is recommended to be used for weld cladding in practice.

  15. Electron Beam Welding of a Depleted Uranium Alloy to Niobium Using a Calibrated Electron Beam Power Density Distribution

    International Nuclear Information System (INIS)

    Elmer, J.W.; Teruya, A.T.; Terrill, P.E.

    2000-01-01

    Electron beam test welds were made joining flat plates of commercially pure niobium to a uranium-6wt%Nb (binary) alloy. The welding parameters and joint design were specifically developed to minimize mixing of the niobium with the U-6%Nb alloy. A Modified Faraday Cup (MFC) technique using computer-assisted tomography was employed to determine the precise power distribution of the electron beam so that the welding parameters could be directly transferred to other welding machines and/or to other facilities

  16. Uranium

    International Nuclear Information System (INIS)

    Whillans, R.T.

    1981-01-01

    Events in the Canadian uranium industry during 1980 are reviewed. Mine and mill expansions and exploration activity are described, as well as changes in governmental policy. Although demand for uranium is weak at the moment, the industry feels optimistic about the future. (LL)

  17. Report of the panel on the use of depleted uranium alloys for large caliber long rod kinetic energy penetrators

    International Nuclear Information System (INIS)

    Sandstrom, D.J.; Jessen, N.; Loewenstein, P.; Weirick, L.

    1980-01-01

    In early 1977 the National Materials Advisory Board, an operating unit in the Commission on Sociotechnical Systems of the National Research Council, NAS/NAE, formed a study committee on High Density Materials for Kinetic Energy Penetrators. The Specific objectives of the Committee were defined as follows. Assess the potential of two materials for use in kinetic energy penetrators, including such factors as: (a) properties (as applied to this application: strength, toughness, and dynamic behavior); (b) uniformity, reliability and reproducibility; (c) deterioration in storage; (d) production capability; (e) ecological impact; (f) quality assurance; (g) availability, and (h) cost. The Committee was divided into two Panels; one panel devoted to the study of tungsten alloys and the other devoted to the study of depleted uranium alloys for use in Kinetic energy penetrators. This report represents the findings and recommendation of the Panel on Uranium

  18. Uranium-Based Cermet Alloys; Cermets a base d'uranium; Metallokeramicheskie splavy na osnove urana; Cermets a base de uranio

    Energy Technology Data Exchange (ETDEWEB)

    Ivanov, V. E.; Zelenskij, V. F.; Voloshchuk, A. I.; Grishok, V. N. [Fiziko-Tekhnicheskij Institut an USSR, Khar' kov, SSSR (Russian Federation)

    1963-11-15

    The paper describes certain features of dispersion-hardened uranium-based cermets. As possible hardening materials, consideration was given to UO{sub 2}, UC, Al{sub 2}O{sub 3}, MgO and UBe{sub 13}. Data were obtained on the behaviour of uranium alloys containing the above-mentioned admixtures during creep tests, short-term strength tests and cyclic thermal treatment. The corrosion resistance o f UBe{sub 13}-based uranium alloys was also studied. )author) [French] Les auteurs decrivent certaines proprietes de cermets a base d'uranium, dont la resistance a ete accrue a l'aide de particules dispersees. Les materiaux utilises a cette fin sont notamment: UO{sub 2}, UC, Al{sub 2}O{sub 3}, MgO et UBe{sub 13}. Les auteurs indiquent les donnees obtenues sur le comportement des cermets a l'uranium; durant les essais de fluage, les essais de resistance a court terme et le traitement thermique cyclique, en mentionnant les substances ajoutees. Ils etudient enfin la resistance a la corrosion des cermets d'uranium et UBe{sub 13}. (author) [Spanish] Los autores describen algunas propiedades de los cermets a base de uranio, reforzados por particulas de diversos compuestos en dispersion. En calidad de posibles materiales de refuerzo, ensayaron el UO{sub 2}, el UC, el Al{sub 2}O{sub 3}, el MgO y el UBe{sub 13}. Obtuvieron datos sobre el comportamiento de esas aleaciones en ensayos de fluencia, ensayoe rapidos de resistencia y tratamiento termico ciclico. Por ultimo, estudiaron la resistencia a la corrosion de las aleaciones de uranio a base de UBe{sub 13}. (author) [Russian] Daetsya opisanie nekotorykh svojstv metallokeramicheskikh splavov urana, uprochnennykh dispersionnymi chastitsami. V kachestve vozmozhnykh uprochnyayushchikh materialov izuchalis' UO{sub 2}, UC, Al{sub 2}O{sub 3} , MgO i UBe{sub 13}. Polucheny dannye o povedenii splavov urana s ukazannymi primesyami pri kripovykh ispytaniyakh, pri kratkovremennykh prochnostnykh ispytaniyakh i pri tsiklicheskoj termoobrabotke

  19. Uranium

    Energy Technology Data Exchange (ETDEWEB)

    Williams, R M

    1976-01-01

    Evidence of expanding markets, improved prices and the short supply of uranium became abundantly clear in 1975, providing the much needed impetus for widespread activity in all phases of uranium operations. Exploration activity that had been at low levels in recent years in Canada was evident in most provinces as well as the Northwest Territories. All producers were in the process of expanding their uranium-producing facilities. Canada's Atomic Energy Control Board (AECB) by year-end had authorized the export of over 73,000 tons of U/sub 3/0/sub 8/ all since September 1974, when the federal government announced its new uranium export guidelines. World production, which had been in the order of 25,000 tons of U/sub 3/0/sub 8/ annually, was expected to reach about 28,000 tons in 1975, principally from increased output in the United States.

  20. Neutron diffraction study of the deformation mechanisms of the uranium-7 wt.% niobium shape memory alloy

    Energy Technology Data Exchange (ETDEWEB)

    Brown, D.W. [Los Alamos National Lab, Los Alamos, NM 87545 (United States)]. E-mail: dbrown@lanl.gov; Bourke, M.A.M. [Los Alamos National Lab, Los Alamos, NM 87545 (United States); Field, R.D. [Los Alamos National Lab, Los Alamos, NM 87545 (United States); Hults, W.L. [Los Alamos National Lab, Los Alamos, NM 87545 (United States); Teter, D.F. [Los Alamos National Lab, Los Alamos, NM 87545 (United States); Thoma, D.J. [Los Alamos National Lab, Los Alamos, NM 87545 (United States); Vogel, S.C. [Los Alamos National Lab, Los Alamos, NM 87545 (United States)

    2006-04-15

    The shape memory effect (SME) has been reported in the uranium-niobium alloy system in the region of the phase diagram surrounding U-6.5 wt.% Nb. In this regime, the material may have either an {alpha}'' monoclinic (U-6 wt.% Nb), or {gamma}{sup 0} tetragonal structure (U-7 wt.% Nb) and is two phase near 6.5 wt.% niobium. In situ neutron diffraction studies during uniaxial compressive loading of U-7 wt.% Nb indicate that strain in the recoverable region is accommodated by both motion of existing twin boundaries within {gamma}{sup 0}-phase and stress-induced phase transformation from the {gamma}{sup 0} to the {alpha}'' structure. The volume fraction of the {gamma}{sup 0}-phase decreases from 100% initially to {approx}26% after 4% total strain and some reversion is observed on release. The initial stress state of the stress-induced {alpha}'' grains will be discussed as well as the load sharing between the two phases.

  1. Mathematic modeling of reactor fuel radiation creep at example of uranium and its alloys

    International Nuclear Information System (INIS)

    Tarasov, V.A.

    2001-01-01

    The model of a radiation creep is explained within the framework of the mechanism of gliding and climbing dislocations based on the conception of a dislocation as not ideal sink for point radiation defects (PRD). The offered model is efficient for installed concentration PRD, considerably exceeding thermally steady state concentration. The gliding of dislocation are describing as due to moving dislocation kinks in Peierl's relief. The climbing of dislocation are describing as due to moving dislocation jogs. The mathematical model for the computer program simulating the offered model of radiation creep is developed. The complex of the computer programs simulating the radiation creep is developed. The computer simulation researches are conducted and the outcomes of a research of a kinetics of a flexible sliding and climbing dislocation interacting to obstacles of a various type (spherical centre of extension, dislocation prismatic loop and their spatially random distributions) for various installed concentration PRD, external loadings and temperatures are represented. The curves of installed rate of a radiation creep from temperature for uranium and its alloys with small additions of molybdenum (from 0,9 to 1,3 %) are obtained

  2. Uranium

    International Nuclear Information System (INIS)

    Perkin, D.J.

    1982-01-01

    Developments in the Australian uranium industry during 1980 are reviewed. Mine production increased markedly to 1841 t U 3 O 8 because of output from the new concentrator at Nabarlek and 1131 t of U 3 O 8 were exported at a nominal value of $37.19/lb. Several new contracts were signed for the sale of yellowcake from Ranger and Nabarlek Mines. Other developments include the decision by the joint venturers in the Olympic Dam Project to sink an exploration shaft and the release of an environmental impact statement for the Honeymoon deposit. Uranium exploration expenditure increased in 1980 and additions were made to Australia's demonstrated economic uranium resources. A world review is included

  3. Uranium

    International Nuclear Information System (INIS)

    Gabelman, J.W.; Chenoweth, W.L.; Ingerson, E.

    1981-01-01

    The uranium production industry is well into its third recession during the nuclear era (since 1945). Exploration is drastically curtailed, and many staffs are being reduced. Historical market price production trends are discussed. A total of 3.07 million acres of land was acquired for exploration; drastic decrease. Surface drilling footage was reduced sharply; an estimated 250 drill rigs were used by the uranium industry during 1980. Land acquisition costs increased 8%. The domestic reserve changes are detailed by cause: exploration, re-evaluation, or production. Two significant discoveries of deposits were made in Mohave County, Arizona. Uranium production during 1980 was 21,850 short tons U 3 O 8 ; an increase of 17% from 1979. Domestic and foreign exploration highlights were given. Major producing areas for the US are San Juan basin, Wyoming basins, Texas coastal plain, Paradox basin, northeastern Washington, Henry Mountains, Utah, central Colorado, and the McDermitt caldera in Nevada and Oregon. 3 figures, 8 tables

  4. Analysis of anisotropic damage in forged Al–Cu–Mg–Si alloy based on creep tests, micrographs of fractured specimen and digital image correlations

    Energy Technology Data Exchange (ETDEWEB)

    Gariboldi, Elisabetta, E-mail: elisabetta.gariboldi@polimi.it [Politecnico di Milano, Department of Mechanical Engineering, Via La Masa 34 20156 Milano (Italy); Naumenko, Konstantin, E-mail: konstantin.naumenko@ovgu.de [Otto-von-Guericke-University Magdeburg, Institute of Mechanics, D-39106 Magdeburg (Germany); Ozhoga-Maslovskaja, Oksana, E-mail: oksana.ozhogamaslovskaja@gmail.com [Politecnico di Milano, Department of Mechanical Engineering, Via La Masa 34 20156 Milano (Italy); Zappa, Emanuele, E-mail: emanuele.zappa@polimi.it [Politecnico di Milano, Department of Mechanical Engineering, Via La Masa 34 20156 Milano (Italy)

    2016-01-15

    The aim of this paper is to analyze anisotropic damage mechanisms in forged Al–Cu–Mg–Si alloy based on the results of creep tests. Smooth specimens are sampled in three forging directions. Creep strain vs. time curves as well as light optical microscope and scanning electron microscope observations illustrate basic features of damage growth. Flat notch specimens are sampled in different directions to analyze stress redistributions and damage in zones of stress concentration. The digital image correlation technique has been applied in situ in order to extract the strain values on the surface of the notched specimens. All observations demonstrate that the principal origins of anisotropic creep and damage are associated with elongated grains and second phase clustered particles located at grain boundaries. Longitudinal specimens possess nucleations of decohesion sites and growth of voids around second phase particles at grain boundaries. Damage evolution for radial and transverse specimens is due to the formation and growth of cracks in second phase particles orthogonal to the principal stress axis. Residual strains are confined to the notch root as well as to the flanges of advanced macrocrack, indicating the small scale yielding during the creep fracture process.

  5. Uranium

    International Nuclear Information System (INIS)

    Anon.

    1983-01-01

    Recent decisions by the Australian Government will ensure a significant expansion of the uranium industry. Development at Roxby Downs may proceed and Ranger may fulfil two new contracts but the decision specifies that apart from Roxby Downs, no new mines should be approved. The ACTU maintains an anti-uranium policy but reaction to the decision from the trade union movement has been muted. The Australian Science and Technology Council (ASTEC) has been asked by the Government to conduct an inquiry into a number of issues relating to Australia's role in the nuclear fuel cycle. The inquiry will examine in particular Australia's nuclear safeguards arrangements and the adequacy of existing waste management technology. In two additional decisions the Government has dissociated itself from a study into the feasibility of establishing an enrichment operation and has abolished the Uranium Advisory Council. Although Australian reserves account for 20% of the total in the Western World, Australia accounts for a relatively minor proportion of the world's uranium production

  6. Uranium

    International Nuclear Information System (INIS)

    Anon.

    1981-01-01

    The French Government has decided to freeze a substantial part of its nuclear power programme. Work has been halted on 18 reactors. This power programme is discussed, as well as the effect it has on the supply of uranium by South Africa

  7. Major constituent quantitative determination in uranium alloys by coupled plasma atomic emission spectrometry and X ray fluorescence wavelength dispersive spectrometry

    International Nuclear Information System (INIS)

    Oliveira, Luis Claudio de; Silva, Adriana Mascarenhas Martins da; Gomide, Ricardo Goncalves; Silva, Ieda de Souza

    2013-01-01

    A wavelength-dispersive X-ray fluorescence (WD-XRF) spectrometric method for determination of major constituents elements (Zr, Nb, Mo) in Uranium/Zirconium/Niobium and Uranium/Molybdenum alloy samples were developed. The methods use samples taken in the form of chips that were dissolved in hot nitric acid and precipitate particles melted with lithium tetraborate and dissolved in hot nitric acid and finally analyzed as a solution. Studies on the determination by inductively coupled plasma optic emission spectrometry (ICP OES) using matched matrix in calibration curve were developed. The same samples solution were analyzed in both methods. The limits of detection (LOD), linearity of the calibrations curves, recovery study, accuracy and precision of the both techniques were carried out. The results were compared. (author)

  8. A survey of the mechanical properties of uranium alloys U-5Mo-3Nb wt.% and U-3Mo-3Nb wt.%

    Energy Technology Data Exchange (ETDEWEB)

    Dupont, G.

    1969-04-15

    In a continuing program on the development of soft and ductile uranium alloys for armament applications, two compositions were studied. These gamma extruded uranium alloys were U-5Mo-3Nb wt.% and U-3Mo-3Nb wt.%. This study was carried out to determine the influence of tempering heat treatments associated with extrusion on the ductility of these uranium alloys. The mechanical properties of both alloys were measured in the extruded condition, in the extruded and annealed condition and in the quenched and tempered condition. A maximum elongation of 13.7% in tension with a low amount of work hardening was obtained for the U-3Mo-3Nb wt.% alloy after 1 1/2 hours anneal at 1200 deg F (650 deg C) followed by a rapid cooling in water at 70 deg F (21 deg C). A maximum elongation of 17.3% with a large amount of work hardening was obtained for alloy U-5Mo-3Nb wt.% after vacuum annealing, normalizing, gamma phase solubilizing at 1500 deg F (815 deg C) and quenching in water at 700 deg F (210 deg C). The maximum ductility achieved in these two alloys by our approaches is low compared with the ductility of Armco Iron employed for the same applications in the field of ballistics.

  9. A study of probabilistic fatigue crack propagation models in Mg Al Zn alloys under different specimen thickness conditions by using the residual of a random variable

    International Nuclear Information System (INIS)

    Choi, Seon Soon

    2012-01-01

    The primary aim of this paper was to evaluate several probabilistic fatigue crack propagation models using the residual of a random variable, and to present the model fit for probabilistic fatigue behavior in Mg Al Zn alloys. The proposed probabilistic models are the probabilistic Paris Erdogan model, probabilistic Walker model, probabilistic Forman model, and probabilistic modified Forman models. These models were prepared by applying a random variable to the empirical fatigue crack propagation models with these names. The best models for describing fatigue crack propagation models with these names. The best models for describing fatigue crack propagation models with these names. The best models for describing fatigue crack propagation models with these names. The best models vor describing fatigue crack propagation behavior in Mg Al Zn alloys were generally the probabilistic Paris Erdogan and probabilistic Walker models. The probabilistic Forman model was a good model only for a specimen with a thickness of 9.45mm

  10. Determination of five kinds of impurity elements such as titanium in uranium titanium alloy by ICP-OES

    International Nuclear Information System (INIS)

    Jiao Yan; Hu Haihong

    2010-01-01

    New description is given of an ICP-OES method in which 5 impurities, Ti, Fe, Ni, Cu, and Al in U-Ti alloy can be determined simultaneously. Studying the dissolution of the sample preparation, separation condition of impurity elements; determining analysis of instrument line, detection limit and detection lower limit; eliminating the matrix effect of Ti and TiO 2 on the measurement of precipitation; standard addition method verify the method accuracy and precision. The results show: taking Uranium titanium alloys containing 0.1000 g sample, 5 kinds of elements Ti detection lower limits is 0.2-0.7 μg·g -1 , recovery were in the range of 98.8%-102.1%, and RSD'S found were less than 8%. The method of measurement proved is accurate and reliable. (authors)

  11. The effect of cooling rate from the γ-phase on the strain-rate sensitivity of a uranium 2 sup(w)/o molybdenum alloy

    International Nuclear Information System (INIS)

    Boyd, G.A.C.; Harding, J.

    1983-01-01

    Tensile tests have been performed at strain rates from 10 -4 to about 2000/s and temperatures from -150 deg C to +250 deg C on a uranium 2 w/o molybdenum alloy which had been aged for 2 hours at 500 deg C after a fast gas cool from the γ-phase at a controlled rate of 40 deg C/minute. The results are compared with those for standard as-extruded material which had received the same aging treatment. Stress-strain curves are presented and the effect of strain rate and temperature on the flow stress, the ultimate tensile stress and the elongation to fracture is determined. A thorough structural characterisation of the specimen materials, using X-ray analysis and scanning and transmission electron microscopy, allows the different mechanical responses to be related to the corresponding microstructural state of the material. Flow stress data at different temperatures and strain rates are analysed in terms of the theory of thermally-activated flow and estimates made of the various activation parameters. (author)

  12. The Static and Fatigue Behavior of AlSiMg Alloy Plain, Notched, and Diamond Lattice Specimens Fabricated by Laser Powder Bed Fusion

    Directory of Open Access Journals (Sweden)

    Hugo Soul

    2018-04-01

    Full Text Available The fabrication of engineered lattice structures has recently gained momentum due to the development of novel additive manufacturing techniques. Interest in lattice structures resides not only in the possibility of obtaining efficient lightweight materials, but also in the functionality of pre-designed architectured structures for specific applications, such as biomimetic implants, chemical catalyzers, and heat transfer devices. The mechanical behaviour of lattice structures depends not only the composition of the base material, but also on the type and size of the unit cells, as well as on the material microstructure resulting from a specific fabrication procedure. The present work focuses on the static and fatigue behavior of diamond cell lattice structures fabricated from an AlSiMg alloy by laser powder bed fusion technology. In particular, the specimens were fabricated with three different orientations of lattice cells—[001], [011], [111]—and subjected to static tensile testing and force-controlled pull–pull fatigue testing up to 1 × 107 cycles. In parallel, the mechanical behavior of dense tensile plain and notched specimens was also studied and compared to that of their lattice counterparts. Results showed a significant effect of the cell orientation on the fatigue lives: specimens oriented at [001] were ~30% more fatigue-resistant than specimens oriented at [011] and [111].

  13. Critical assessment of precracked specimen configuration and experimental test variables for stress corrosion testing of 7075-T6 aluminum alloy plate

    Science.gov (United States)

    Domack, M. S.

    1985-01-01

    A research program was conducted to critically assess the effects of precracked specimen configuration, stress intensity solutions, compliance relationships and other experimental test variables for stress corrosion testing of 7075-T6 aluminum alloy plate. Modified compact and double beam wedge-loaded specimens were tested and analyzed to determine the threshold stress intensity factor and stress corrosion crack growth rate. Stress intensity solutions and experimentally determined compliance relationships were developed and compared with other solutions available in the literature. Crack growth data suggests that more effective crack length measurement techniques are necessary to better characterize stress corrosion crack growth. Final load determined by specimen reloading and by compliance did not correlate well, and was considered a major source of interlaboratory variability. Test duration must be determined systematically, accounting for crack length measurement resolution, time for crack arrest, and experimental interferences. This work was conducted as part of a round robin program sponsored by ASTM committees G1.06 and E24.04 to develop a standard test method for stress corrosion testing using precracked specimens.

  14. Interdiffusion, Intrinsic Diffusion, Atomic Mobility, and Vacancy Wind Effect in γ(bcc) Uranium-Molybdenum Alloy

    Science.gov (United States)

    Huang, Ke; Keiser, Dennis D.; Sohn, Yongho

    2013-02-01

    U-Mo alloys are being developed as low enrichment uranium fuels under the Reduced Enrichment for Research and Test Reactor (RERTR) Program. In order to understand the fundamental diffusion behavior of this system, solid-to-solid pure U vs Mo diffusion couples were assembled and annealed at 923 K, 973 K, 1073 K, 1173 K, and 1273 K (650 °C, 700 °C, 800 °C, 900 °C, and 1000 °C) for various times. The interdiffusion microstructures and concentration profiles were examined via scanning electron microscopy and electron probe microanalysis, respectively. As the Mo concentration increased from 2 to 26 at. pct, the interdiffusion coefficient decreased, while the activation energy increased. A Kirkendall marker plane was clearly identified in each diffusion couple and utilized to determine intrinsic diffusion coefficients. Uranium intrinsically diffused 5-10 times faster than Mo. Molar excess Gibbs free energy of U-Mo alloy was applied to calculate the thermodynamic factor using ideal, regular, and subregular solution models. Based on the intrinsic diffusion coefficients and thermodynamic factors, Manning's formalism was used to calculate the tracer diffusion coefficients, atomic mobilities, and vacancy wind parameters of U and Mo at the marker composition. The tracer diffusion coefficients and atomic mobilities of U were about five times larger than those of Mo, and the vacancy wind effect increased the intrinsic flux of U by approximately 30 pct.

  15. Low content uranium alloys for nuclear fuels; Alliages d'uranium a faible teneur pour elements combustibles

    Energy Technology Data Exchange (ETDEWEB)

    Aubert, H.; Laniesse, J. [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1964-07-01

    A description is given of the structure and the properties of low content alloys containing from 0.1 to 0.5 per cent by weight of Al, Fe, Cr, Si, Mo or a combination of these elements. A study of the kinetics and of the mode of transformation has made it possible to choose the most satisfactory thermal treatment. An attempt has been made to prepare alloys suitable for an economical industrial development having a small {alpha} grain structure without marked preferential orientation, with very fine and stable precipitates as well as a high creep-resistance. The physical properties and the mechanical strength of these alloys are given for temperatures of 20 to 600 deg C. These alloys proved very satisfactory when irradiated in the form of normal size fuel elements. (authors) [French] Sont decrits la structure et les proprietes d'alliages a faible teneur, contenant de 0,1 a 0,5 pour cent en poids de Al, Fe, Cr, Si, Mo ou une combinaison de ces elements. L'etude des cinetiques et du mode de transformation permet de choisir le traitement thermique le plus favorable. On a cherche a mettre, au point des alliages se pretant a une mise en oeuvre industrielle economique et presentant une structure a petits grains {alpha}, sans orientation preferentielle marquee, avec des precipites tres fins et stables ainsi qu'une bonne resistance au fluage. Les proprietes physiques et la resistance mecanique de ces alliages sont decrites entre la temperature ambiante et 600 deg C. Irradies sous forme d'elements combustibles de dimensions normales, ces alliages ont montre un bon comportement. (auteurs)

  16. Low content uranium alloys for nuclear fuels; Alliages d'uranium a faible teneur pour elements combustibles

    Energy Technology Data Exchange (ETDEWEB)

    Aubert, H; Laniesse, J [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1964-07-01

    A description is given of the structure and the properties of low content alloys containing from 0.1 to 0.5 per cent by weight of Al, Fe, Cr, Si, Mo or a combination of these elements. A study of the kinetics and of the mode of transformation has made it possible to choose the most satisfactory thermal treatment. An attempt has been made to prepare alloys suitable for an economical industrial development having a small {alpha} grain structure without marked preferential orientation, with very fine and stable precipitates as well as a high creep-resistance. The physical properties and the mechanical strength of these alloys are given for temperatures of 20 to 600 deg C. These alloys proved very satisfactory when irradiated in the form of normal size fuel elements. (authors) [French] Sont decrits la structure et les proprietes d'alliages a faible teneur, contenant de 0,1 a 0,5 pour cent en poids de Al, Fe, Cr, Si, Mo ou une combinaison de ces elements. L'etude des cinetiques et du mode de transformation permet de choisir le traitement thermique le plus favorable. On a cherche a mettre, au point des alliages se pretant a une mise en oeuvre industrielle economique et presentant une structure a petits grains {alpha}, sans orientation preferentielle marquee, avec des precipites tres fins et stables ainsi qu'une bonne resistance au fluage. Les proprietes physiques et la resistance mecanique de ces alliages sont decrites entre la temperature ambiante et 600 deg C. Irradies sous forme d'elements combustibles de dimensions normales, ces alliages ont montre un bon comportement. (auteurs)

  17. Annex 4 - Task 08/13 final report, Producing the binary uranium alloys with alloying components Al, Mo, Zr, Nb, and B; Prilog 4 - Zavrsni izvestaj o podzadatku 08/13, Dobijanje binarnih legura urana sa legirajucim komponentama Al, Mo, Zr, Nb i B

    Energy Technology Data Exchange (ETDEWEB)

    Lazarevic, Dj [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1961-12-15

    Due to reactivity of uranium in contact with the gasses O{sub 2}, N{sub 2}, H{sub 2}, especially under higher temperatures uranium processing is always done in vacuum or inert gas. Melting, alloying and casting is done in high vacuum stoves. This report reviews the type of furnaces and includes detailed description of the electric furnace for producing uranium alloys which is available in the Institute.

  18. Uranium

    International Nuclear Information System (INIS)

    Battey, G.C.; McKay, A.D.

    1988-01-01

    Production for 1986 was 4899 t U 3 O 8 (4154 t U), 30% greater than in 1985, mainly because of a 39% increase in production at Ranger. Exports for 1986 were 4166 t U 3 O 8 at an average f.o.b. unit value of $40.57/lb U 3 O 8 . Private exploration expenditure for uranium in Australia during the 1985-86 fiscal year was $50.2 million. Plans were announced to increase the nominal capacity of the processing plant at Ranger from 3000 t/year U 3 O 8 to 4500 t and later to 6000 t/year. Construction and initial mine development at Olympic Dam began in March. Production is planned for mid 1988 at an annual rate of 2000 t U 3 O 8 , 30 000 t Cu, and 90 000 oz (2800 kg) Au. The first long-term sales agreement was concluded in September 1986. At the Manyingee deposit, testing of the alkaline solution mining method was completed, and the treatment plant was dismantled. Spot market prices (in US$/lb U 3 O 8 ) quoted by Nuexco were generally stable. From January-October the exchange value fluctuated from US$17.00-US$17.25; for November and December it was US$16.75. Australia's Reasonably Assured Resources of uranium recoverable at less than US$80/kg U at December 1986 were estimated as 462 000 t U, 3000 t U less than in 1985. This represents 30% of the total low-cost RAR in the WOCA (World Outside the Centrally Planned Economy Areas) countries. Australia also has 257 000 t U in the low-cost Estimated Additional Resources Category I, 29% of the WOCA countries' total resources in this category

  19. Properties of low content uranium-molybdenum alloys which may be used as nuclear fuels; Proprietes des alliages uranium-molybdene de faibles teneurs utilisables comme materiaux combustibles

    Energy Technology Data Exchange (ETDEWEB)

    Lehmann, J; Decours, J [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1964-07-01

    Metallurgical properties are given in this report of uranium-molybdenum alloys containing 0,5 to 3 per cent of molybdenum. Since some of these alloys are used in EDF power reactors are given: briefly the operating conditions imposed on nuclear fuels: maximum temperature, temperature gradient and external pressure. In the first part are considered the structural properties of the alloys correlation with the phase transformation kinetics; a description is given of the effects of certain physico-metallurgical factors on the morphology and the crystalline structure of the materials: - solidification conditions and the heredity of the {gamma} structure, - cooling rate at the transformation points, - whether or not the intermediate {gamma} {yields} {beta} transformation is suppressed In the second part we show how a knowledge of the phase transformation processes has made it possible to define the optimum preparation conditions for these materials in the form of fuel tubes intended for the EDF reactors: casting conditions, controlled cooling treatments, weldability. In the third part we study the thermal, stability during the long duration high temperature treatments and the cycles in the two zones of the diagram {alpha} + {gamma}; {beta} + {gamma} the effects of the morphology (in particular the two types of {alpha} pseudo-grains observed) and of the cooling rate during the transformation point transitions are described. In the fourth part are discussed the mechanical properties: resistance to a tractive force, resistance to creep, resilience. These properties can also be affected by the {gamma} structure heredity and by the cooling rate to which the alloy has been subjected. In conclusion we discuss the reasons which led to the choice of some of these alloys for the first EDF reactors in particular the advantages of their high creep resistance between 450 and 600 deg C for use in the form of tubes subjected to an external pressure. (authors) [French] Dans ce rapport

  20. Evaluation of methods for cleaning low carbon uranium metal and alloy samples

    International Nuclear Information System (INIS)

    Kirchner, K.; Dixon, M.

    1979-01-01

    Several methods for cleaning uranium samples prior to carbon analysis, using a Leco Carbon Analyzer, were evaluated. Use of Oakite Aluminum NST Cleaner followed by water and acetone rinse was found to be the best overall technique

  1. Radiation damage of uranium

    International Nuclear Information System (INIS)

    Lazarevic, Dj.

    1966-11-01

    Study of radiation damage covered the following: Kinetics of electric resistance of uranium and uranium alloy with 1% of molybdenum dependent on the second phase and burnup rate; Study of gas precipitation and diffusion of bubbles by transmission electron microscopy; Numerical analysis of the influence of defects distribution and concentration on the rare gas precipitation in uranium; study of thermal sedimentation of uranium alloy with molybdenum; diffusion of rare gas in metal by gas chromatography method

  2. Numerical Analysis of Crack Progress in Different Areas of a Friction Stir Welded Bead for an 5251 H14 Aluminum Alloy Specimen

    Directory of Open Access Journals (Sweden)

    Y. Kambouz

    2014-02-01

    Full Text Available The assemblies welded by Friction Stir Welding have a major advantage which is the absence of a metal filler. This process contributes to the welding of materials that are known to be difficult to weld using the conventional techniques often employed in the field of transport, for example in the automobile body by applying a spot welding. The numerical modeling of this type of process is complex, not only in terms of the variety of physical phenomena which must be considered, but also because of the experimental procedure that must be followed in order to verify and validate numerical predictions. In this work, a finite element model is proposed in order to simulate the crack propagation under monotonic loading in different areas of the weld seam of a strain hardening CT-50 aluminum alloy 5251H14 specimen.

  3. Development of an aging integrator for uranium-0.75 weight percent titanium alloy part aging control

    International Nuclear Information System (INIS)

    Howington, L.C.

    1977-12-01

    An instrumentation system (Aging Integrator) has been developed to provide more precise control of the heat-treatment process used on uranium-0.75 wt.% titanium alloy material. The Aging Integrator calculates the integral of a predetermined aging function to control the aging period in the heat-treatment process. This control was employed to compensate for discrepancies caused by variations in heatup times, furnace-control fluctuations, and disagreement as to the temperature at which aging actually starts. Although the Aging Integrator hardware has been installed and satisfactorily tested on a production-area furnace, sufficient data to estimate a statistically sound aging integration function will not be available for approximately one year

  4. Study of uranium-plutonium alloys containing from 0 to 20 peri cent of plutonium (1963); Etude des alliages uranium-plutonium aux concentrations comprises entre 0 et 20 pour cent de plutonium (1963)

    Energy Technology Data Exchange (ETDEWEB)

    Paruz, H [Commissariat a l' Energie Atomique, Fontenay-aux-Roses (France). Centre d' Etudes Nucleaires

    1963-05-15

    The work is carried out on U-Pu alloys in the region of the solid solution uranium alpha and in the two-phase region uranium alpha + the zeta phase. The results obtained concern mainly the influence of the addition of plutonium on the physical properties of the uranium (changes in the crystalline parameters, the density, the hardness) in the region of solid solution uranium alpha. In view of the discrepancies between various published results as far as the equilibrium diagram for the system U-Pu is concerned, an attempt was made to verify the extent of the different regions of the phase diagram, in particular the two phased-region. Examinations carried out on samples after various thermal treatments (in particular quenching from the epsilon phase and prolonged annealings, as well as a slow cooling from the epsilon phase) confirm the results obtained at Los Alamos and Harwell. (author) [French] L'etude porte sur des alliages U-Pu du domaine de la solution solide uranium alpha et du domaine biphase uranium + phase zeta. Les resultats obtenus concernent en premier lieu l'influence de l'addition de plutonium sur les proprietes physiques de l'uranium (changement des parametres cristallins, densite, durete) dans le domaine de la solution solide uranium alpha. Compte tenu des divergences entre les differents resultats publies en ce qui concerne le diagramme d'equilibre du systeme U-Pu, on a essaye ensuite de verifier l'etendue des differents domaines du diagramme des phases, en particulier du domaine biphase zeta + uranium alpha. Les examens par micrographie et par diffraction des rayons X des echantillons apres differents traitements thermiques (notamment trempe a partir de la phase epsilon et recuits prolonges, ainsi qu'un refroidissement lent etage a partir de la phase epsilon) confirment les resultats obtenus a Los Alamos et a Harwell. (auteur)

  5. 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

  6. Mechanical properties of aluminium-uranium alloy and aluminium commercially pure at several temperatures

    International Nuclear Information System (INIS)

    Quadros, N.F. de.

    1976-01-01

    The mechanical properties of Ai-U (18,4 wt %) alloy with and without heat treatment were determined, and they were compared with the mechanical properties of aluminum alloy of commercial purity, AI-1100, at tempiratures of 25, 500, 550 and 600 0 C, the changes of both the yield point stress and the ultimate tensile strength as a function of temperature may be described through two emperical relationships. A fractography study was also made [pt

  7. Uranium and plutonium extraction from fluoride melts by lithium-tin alloys

    International Nuclear Information System (INIS)

    Kashcheev, I.N.; Novoselov, G.P.; Zolotarev, A.B.

    1975-01-01

    Extraction of small amounts of uranium (12 wt. % concentration) and plutonium (less than 1.10sup(-10) % concentration) from lithium fluoride melts into the lithium-tin melts is studied. At an increase of temperature from 850 to 1150 deg the rate of process increases 2.5 times. At an increase of melting time the extraction rapidly enhances at the starting moment and than its rate reduces. Plutonium is extracted into the metallic phase for 120 min. (87-96%). It behaves analogously to uranium

  8. Thermodynamic study contribution of U-Fe and U-Ga alloys by high temperature mass spectroscopy, and of the wetting of yttrium oxide by uranium

    International Nuclear Information System (INIS)

    Gardie, P.

    1992-01-01

    High temperature thermodynamic properties study of U-Fe and U-Ga alloys, and wetting study of yttrium oxide by uranium are presented. High temperature mass spectrometry coupled to a Knudsen effusion multi-cell allows to measure iron activity in U-Fe alloys and of gallium in U-Ga alloys, the U activity is deduced from Gibbs-Duhem equation. Wetting of the system U/Y_2O_3_-_x is studied between 1413 K and 1973 K by the put drop method visualized by X-rays. This technique also furnishes density, surface tension of U and of U-Fe alloys put on Y_2O_3_-_x. A new model of the interfacial oxygen action on wetting is done for the system U/Y_2O_3_-_x. (A.B.)

  9. Review of DREV uranium research

    International Nuclear Information System (INIS)

    Drolet, J.P.; Erickson, W.H.; Tardif, H.P.

    1976-01-01

    This report presents a brief review of the DREV uranium research carried out on various aspects of the physical metallurgy of depleted uranium alloys. It includes (1) a survey of the early work on polynary alloys, (2) recent metallurgical investigations on various alloy systems and (3) miscellaneous studies on grain size refinement, grain growth, powder metallurgy, pyrophoricity and directional casting of uranium alloys. A general summary of most of the studies carried out during the last ten years is also presented

  10. Study of the transformation of uranium-niobium alloys with low niobium concentrations, tempered from the gamma and beta + gamma 1 regions and then annealed at different temperatures. Comparison with uranium-molybdenum alloys (1963); Etude des transformations des alliages uranium-niobium a faible teneur en niobium trempes depuis les domaines gamma et beta + gamma 1 puis revenus a differentes temperatures. Comparaison avec les alliages uranium-molybdene (1963)

    Energy Technology Data Exchange (ETDEWEB)

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

    1962-09-15

    The author shows that uranium-niobium alloys, like uranium-molybdenum alloys, tempered from the gamma region, give a martensitic phase with a structure deriving from that of alpha uranium by a slight contraction parallel to the axis [001], The critical cooling rate allowing the formation of this martensite is 80 deg. C/s at 750 deg. C. Retention of the beta phase of uranium-niobium alloys is particularly difficult, the critical retention rate being 700 deg. C/s at 668 deg. C for an alloy containing 2.5 at. per cent of Nb. This beta phase is completely converted to the alpha phase at room temperature in about 6 hours. The TTT curves of this beta alloy are effectively reduced to the lower branch of the lower 'C'. The beta phase conversion law is expressed as: 1-x = exp. (kt){sup n} x being the degree of progression of the conversion, t the time, n an exponent no-varying with temperature and having approximately the value 2 for the alloy considered, k an increasing function of temperature. The activation energy of conversion is of the order of 14,600 cal/mole. Niobium is much less active than molybdenum as a stabiliser of beta uranium. (author) [French] Dans ce travail l'auteur montre que les alliages uranium-niobium, comme d'ailleurs les alliages uranium-molybdene, trempes depuis le domaine gamma, donnent une phase martensitique dont la structure derive de celle de l'uranium alpha par une legere contraction parallele de l'axe [001]. La vitesse critique de refroidissement permettant la formation de cette martensite est de 80 deg. C/s a 750 deg. C. La retention de la phase beta des alliages uranium-niobium est particulierement delicate car la vitesse critique de retention est de 700 deg. C/s a 668 deg. C pour l'alliage a 2,5 at. pour cent de Nb. Cette phase beta se transforme completement en phase alpha a la temperature ordinaire en 6 heures environ. Les courbes TTT de cet alliage de structure beta se reduisent pratiquement a la branche inferieure du 'C' inferieur. La

  11. Investigation of point defects diffusion in bcc uranium and U–Mo alloys

    International Nuclear Information System (INIS)

    Smirnova, D.E.; Kuksin, A.Yu.; Starikov, S.V.

    2015-01-01

    We present results of investigation of point defects formation and diffusion in pure γ-U and γ-U–Mo fuel alloys. The study was performed using molecular dynamics simulation with the different interatomic potentials. The point defects formation and migration energies were estimated for bcc γ-U and U–9 wt.%Mo alloy. The calculated diffusivities of atoms via defects are provided for pure γ-U and for the alloy components. Analysis of simulation results shows that self-interstitial atoms play a leading role in the self-diffusion processes in the materials studied. This fact can explain a remarkably high self-diffusion mobility observed experimentally for γ-U. The self-diffusion coefficients in γ-U calculated in this assumption agree with the data measured experimentally. It is shown that alloying of γ-U with Mo increase formation energy for self-interstitial atoms and decelerate their mobility. These changes lead to decrease of self-diffusion coefficients in U–Mo alloy compared to pure U

  12. Modeling of uranium alloy response in plane impact and reverse ballistic experiments

    International Nuclear Information System (INIS)

    Herrmann, B.; Landau, A.; Shvarts, D.; Favorsky, V.; Zaretsky, E.

    2002-01-01

    The dynamic behavior of a solution heat-treated, water-quenched and aged U-0.75wt%Ti alloy was studied in planar (disk-on-disk) and reverse ballistic (disk-on-rod) impact experiments performed with a 25 mm light-gas gun. The impact velocity ranged from 100 to 500 m/sec. The impacted samples were softly recovered for further metallographic examination. The VISAR records of the sample free surface velocity, obtained in planar impact experiments, were simulated with 1-D hydrocode for calibrating the parameters of modified Steinberg-Cochran-Guinan (SCG) constitutive equation of the alloy. The same SCG equation was employed in 2-D AUTODYN simulation of the alloy response in the reverse ballistic experiments, with VISAR monitoring of the lateral sample surface velocity. Varying the parameters of the strain-dependent failure model allows relating the features of the recorded velocity profiles with the results of the examination of the damaged samples

  13. Mechanisms of the plastic deformation of uranium alloys at low temperature

    International Nuclear Information System (INIS)

    Le Poac, P.; Nomine, A.M.; Miannay, D.

    1976-01-01

    The mechanical characteristics of the bcc binary alloys U-6Mo, U-8Mo, U-10Mo, U-12Mo and bcc ternary alloys U-8Mo-1Ti, U-10Mo-1Ti, U-10Mo-1Zr, stressed in compression, were determined between -196 deg C and + 450 deg C. The plastic flow shear stress in non-dependent on temperature above 300 deg C. At lower temperature shear stress is highly activated, except for the alloy U-6Mo and U-12Mo. Athermal shear stress above 300 deg C is due to the hardening of the solid solution described by Mott and Nabarro. In the thermal range, the recombination of the dissociated dislocations controls the plastic deformation [fr

  14. Influence of the alloying elements vanadium, chromium and carbon on the electrochemical behavior of uranium in media with a pH 13 or a pH acid

    International Nuclear Information System (INIS)

    Pommier, Gerard; Jouve, Gerard; Lacombe, Paul.

    1976-06-01

    The electrochemical properties of uranium alloys with low vanadium and chromium contents were studied in aqueous medium for different pH values of the solution (pH between 0 and 5 in H 2 SO 4 medium and pH=13 in NaOH medium). In acid medium, the study of the behavior of the two types of alloys carried out by the potentiokinetic method is described. The specific role of chromium concerning the anodic process is demonstrated and the influence of vanadium in specimens of same nominal vanadium contents but different carbon contents is revealed by the modification of the reduction overvoltage of water. In basic medium, the electrochemical study was supported by an optical method of determining the relative growth kinetics of the films in situ and continuously. At lower values of potential, the growth of an oxide film of UO 2 with linear growth kinetics is demonstrated; at higher values of potential a system of two layers is observed and its evolution is followed kinematically. The film initially formed is constituted of an oxide UO 3 2H 2 O, and its growth is linear, then a film of UO 2 develops underneath. A structural evolution of the superficial film is then observed, an evolution which leads to its cracking after breakdown. These phenomena were followed by electron microscopy using a technic of two stage replicas [fr

  15. Metallic uranium as fuel for fast reactors

    International Nuclear Information System (INIS)

    Moura Neto, C. de

    1988-01-01

    This paper presents a first overview of the use of metallic uranium and its alloys as an option for fuel for rapid reactors. Aspects are discussed concerning uranium alloys which present high solubility in the gamma phase. (author)

  16. Contribution towards the study of {beta}{yields}{alpha} transformation in uranium and its alloys (1962); Contribution a l'etude de la transformation {beta}{yields}{alpha} dans l'uranium et ses alliages (1962)

    Energy Technology Data Exchange (ETDEWEB)

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

    1962-05-15

    The kinetics of the transformation of uranium alloys containing 0.5 - 0.75 - 1.0 - 1.5 and 3 atoms per cent have been studied. The influence of heat treatment before decomposition has been discussed. The study of the transformation characteristics such as kinetics, residual phases, phenomena connected with the coherence between phases, reversibility below the equilibrium temperature, shows the following mechanisms exhibited during the decomposition of the {beta} phase on lowering the temperature: 1 ) eutectoid, 2) bainitic, 3) martensitic. The study of the TTT diagrams of alloys containing decreasing percentages of chromium indicates that the unalloyed uranium transforms without maintaining the coherence above 600 deg. C, where as at lower temperatures the transformation is mainly martensitic. The various alloying elements can be characterised by their influence on the three TTT curves corresponding to the three possible transformation mechanisms. The ability of the uranium alloys to alpha grain refining during isothermal decomposition or ambient temperature quenching is directly connected with the characteristics of the TTT diagrams and especially to the mode of bainitic transformation. (author) [French] II a ete etudie la cinetique de transformation des alliages uranium-chrome de teneur 0,5 - 0,75 - 1 - 1,5 - et 3 atomes pour cent. L'influence des traitements thermiques precedant la decomposition a ete discutee. L'etude des caracteristiques de la transformation: cinetique, phases residuelles, phenomenes lies a la coherence entre phases, reversibilite au-dessous de la temperature d'equilibre, permet de conclure que la decomposition met en jeu successivement les trois mecanismes eutectoide, bainitique et martensitique quand la temperature baisse. L'etude de l'evolution des diagrammes TTT quand la teneur en Cr decroit indique que dans l'uranium non allie la transformation se fait sans maintien de la coherence au-dessus de 600 deg. C; a

  17. Contribution towards the study of {beta}{yields}{alpha} transformation in uranium and its alloys (1962); Contribution a l'etude de la transformation {beta}{yields}{alpha} dans l'uranium et ses alliages (1962)

    Energy Technology Data Exchange (ETDEWEB)

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

    1962-05-15

    The kinetics of the transformation of uranium alloys containing 0.5 - 0.75 - 1.0 - 1.5 and 3 atoms per cent have been studied. The influence of heat treatment before decomposition has been discussed. The study of the transformation characteristics such as kinetics, residual phases, phenomena connected with the coherence between phases, reversibility below the equilibrium temperature, shows the following mechanisms exhibited during the decomposition of the {beta} phase on lowering the temperature: 1 ) eutectoid, 2) bainitic, 3) martensitic. The study of the TTT diagrams of alloys containing decreasing percentages of chromium indicates that the unalloyed uranium transforms without maintaining the coherence above 600 deg. C, where as at lower temperatures the transformation is mainly martensitic. The various alloying elements can be characterised by their influence on the three TTT curves corresponding to the three possible transformation mechanisms. The ability of the uranium alloys to alpha grain refining during isothermal decomposition or ambient temperature quenching is directly connected with the characteristics of the TTT diagrams and especially to the mode of bainitic transformation. (author) [French] II a ete etudie la cinetique de transformation des alliages uranium-chrome de teneur 0,5 - 0,75 - 1 - 1,5 - et 3 atomes pour cent. L'influence des traitements thermiques precedant la decomposition a ete discutee. L'etude des caracteristiques de la transformation: cinetique, phases residuelles, phenomenes lies a la coherence entre phases, reversibilite au-dessous de la temperature d'equilibre, permet de conclure que la decomposition met en jeu successivement les trois mecanismes eutectoide, bainitique et martensitique quand la temperature baisse. L'etude de l'evolution des diagrammes TTT quand la teneur en Cr decroit indique que dans l'uranium non allie la transformation se fait sans maintien de la coherence au-dessus de 600 deg. C; a plus basse temperature la

  18. The status of uranium-silicon alloy fuel development for the RERTR program

    International Nuclear Information System (INIS)

    Domagala, R.F.; Wiencek, T.C.; Thresh, H.R.; Stahl, D.

    1983-01-01

    As part of the national Reduced Enrichment Research and Test Reactor (RERTR) Program, Argonne National Laboratory (ANL) is engaged in a fuel-alloy development project. The fuel alloys are dispersed in an aluminum matrix and metallurgically roll-bonded within 6061 Al alloy. To date, 'miniplates' with up to 40 vol. fuel alloy have been successfully fabricated. Thirty-one of these plates have been or are being irradiated in the Oak Ridge Reactor (ORR). Three different fuels have been used in the ANL miniplates: U 3 Si (U + 4 wt.% Si), U 3 Si 2 (U + 7.4 wt.% Si), or ''U 3 SiAl'' (U + 3.5 wt.% Si + 1.5 wt.% Al). All three are candidates for permitting higher fuel loadings and thus lower enrichments of 235 U than would be possible with either UAl x or U 3 O 8 , the current fuels for plate-type elements. The enrichment level employed at ANL is ∼19.8%. Continuing effort involves the production of miniplates with up to ∼60 vol. % fuel, the development of a technology for full-size plate fabrication, and post-irradiation examination of miniplates already removed from the ORR. (author)

  19. Dissolution of metallic uranium and its alloys. Part II. Screening study results: Identification of an effective non-thermal uranium dissolution method

    International Nuclear Information System (INIS)

    Laue, C.A.; Gates-Anderson, D.; Fitch, T.E.

    2004-01-01

    Screening experiments were performed to evaluate reagent systems that deactivate pyrophoric, metallic depleted uranium waste streams at ambient temperature. The results presented led to the selection of two systems, which would be investigated further, for the design of the LLNL onsite treatment process of metallic depleted uranium wastes. The two feasible systems are: (a) 7.5 mol/l H 2 SO 4 - 1 mol/l HNO 3 and (b) 3 mol/l HCl - 1 mol/l H 3 PO 4 . The sulfuric acid system dissolves uranium metal completely, while the hydrochloric-phosphoric acid system converts the metal completely into a solid, which might be suitable for direct disposal. Both systems combine oxidation of metallic uranium with complexation of the uranium ions formed to effectively deactivate uranium.s pyrophoricity at ambient temperature. (author)

  20. Stress field determination in an alloy 600 stress corrosion crack specimen; Determination du champ de contraintes dans une eprouvette de corrosion sous contrainte de l`alliage 600

    Energy Technology Data Exchange (ETDEWEB)

    Rassineux, B.; Labbe, T.

    1995-05-01

    In the context of EDF studies on stress corrosion cracking rates in the Alloy 600 steam generators tubes, we studied the influence of strain hardened surface layers on the different stages of cracking for a tensile smooth specimen (TLT). The stress field was notably assessed to try and explain the slow/rapid-propagation change observed beyond the strain hardened layers. The main difficulty is to simulate in a finite element model the inner and outer surfaces of these strain hardened layers, produced by the final manufacturing stages of SG tubes which have not been heat treated. In the model, the strain hardening is introduced by simulating a multi-layer material. Residual stresses are simulated by an equivalent fictitious thermomechanical calculation, realigned with respect to X-ray measurements. The strain hardening introduction method was validated by an analytical calculation giving identical results. Stress field evolution induced by specimen tensile loading were studied using an elastoplastic 2D finite element calculations performed with the Aster Code. The stress profile obtained after load at 660 MPa shows no stress discontinuity at the boundary between the strain hardened layer and the rest of the tube. So we propose that a complementary calculation be performed, taking into account the multi-cracked state of the strain hardened zones by means of a damage variable. In fact, this state could induce stress redistribution in the un-cracked area, which would perhaps provide an explanation of the crack-ground rate change beyond the strain hardened zone. The calculations also evidence the harmful effects of plastic strains on a strain hardened layer due to the initial state of the tube (not heat-treated), to grit blasting or to shot peening. The initial compressive stress condition of this surface layer becomes, after plastic strain, a tensile stress condition. These results are confirmed by laboratory test. (author). 10 refs., 18 figs., 9 tabs., 2 appends.

  1. Aluminum titanate crucible for molten uranium

    International Nuclear Information System (INIS)

    Asbury, J.J.

    1975-01-01

    An improved crucible for molten uranium is described. The crucible or crucible liner is formed of aluminum titanate which essentially eliminates contamination of uranium and uranium alloys during molten states thereof. (U.S.)

  2. Study of uranium - 20 Wt per cent plutonium-niobium alloys (1963)

    International Nuclear Information System (INIS)

    Abgrall, J.; Barthelemy, P.; Boucher, R.

    1963-01-01

    U-Pu-Nb alloys containing 20 wt per cent Pu and 10 - 20 - 30 - 40 - 50 or 60 wt per cent Nb have been studied principally to determine the feasibility of their use as fuel element. The fabrication, casting and homogenisation presented certain difficulties due specially to niobium. The transformation temperatures, thermal expansion coefficients and nature of phases have been determined by thermal analysis, dilatometry, micrography and X Rays diffraction. For similar compositions, U-Pu-Mo and U-Pu-Nb alloys have many common points concerning the presence of zeta phase (up to 40 wt per cent Nb), the coefficients of expansion, the good behaviour during thermal cycling and the good resistance to air oxidation in spite of zeta phase. In consequence, irradiation tests in EL 3 reactor (Saclay) will be carried out in the near future. (authors) [fr

  3. Microsegregation of heat and homogenization treatments in uranium-niobium alloys (U-Nb)

    International Nuclear Information System (INIS)

    Leal, J.F.

    1988-01-01

    In the following sections microsegration results in U-3,6 Wt% Nb and U-6,1 Wt% Nb alloys casted in noconsumable electrode arc furnace are presented. The microsegration is studied qualitatively by optical microscopy and quantitatively by electron microprobe. The degree of homogenization has been measured after 800 and 850 0 C heat treatments in tubular resistive furnace. The microstructures after heat treatments are quantitatively analysed to check effects on the casting structures, mainly the variations in solute along the dendrite arm spacing. Some solidification phenomena are then discussed on reference to theorical models of dendritic solidification, including microstructure and microsegregation. The experimental results are compared to theoretical on basis of initial and residual microsegregation after homogenization treatments. The times required for homogenization of the alloys are also discussed in function of the microsegregation from casting structures and the temperatures of the treatments. (author) [pt

  4. Micro segregation and homogenization treatments of uranium-niobium alloys (U-Nb)

    International Nuclear Information System (INIS)

    Leal, Jose Fernando

    1988-01-01

    In the following sections micro segregation results in 0-3,6 wt% Nb and U-6,1 wt% Nb alloys casted in no consumable electrode arc furnace are presented. The micro segregation is studied qualitatively by optical microscopy and quantitatively by electron microprobe. The degree of homogenization has been measured after 800 and 850 deg C heat treatments in tubular resistive furnace. The microstructures after heat treatments are quantitatively analysed to check effects on the casting structures, mainly the variations in solute along the dendrite arm spacing. Some solidification phenomena are then discussed on reference to theoretical models of dendritic solidification , including microstructure and micro segregation. The experimental results are compared to theoretical on basis of initial and residual micro segregation after homogenization treatments. The times required for homogenization of the alloys are also discussed in function of the micro segregation from casting structures and the temperatures of the treatments. (author)

  5. Qualification of uranium-molybdenum alloy fuel - conclusions of an international workshop

    International Nuclear Information System (INIS)

    Snelgrove, J.L.; Languille, A.

    2000-01-01

    Thirty-one participants representing 21 reactors, fuel developers, fuel fabricators, and fuel reprocessors in 11 countries discussed the requirements for qualification of U-Mo alloy fuel at a workshop held at Argonne National Laboratory on January 17-18, 2000. Consensus was reached that the qualification plans of the U.S. RERTR program and the French U-Mo fuel development program are valid. The items to be addressed during qualification are summarized in the paper. (author)

  6. Microstructural evolution of a uranium-10 wt.% molybdenum alloy for nuclear reactor fuels

    International Nuclear Information System (INIS)

    Clarke, A.J.; Clarke, K.D.; McCabe, R.J.; Necker, C.T.; Papin, P.A.; Field, R.D.; Kelly, A.M.; Tucker, T.J.; Forsyth, R.T.; Dickerson, P.O.; Foley, J.C.; Swenson, H.; Aikin, R.M.; Dombrowski, D.E.

    2015-01-01

    Low-enriched uranium-10 wt.% molybdenum (LEU-10wt.%Mo) is of interest for the fabrication of monolithic fuels to replace highly-enriched uranium (HEU) dispersion fuels in high performance research and test reactors around the world. In this work, depleted uranium-10 wt.%Mo (DU-10wt.%Mo) is used to simulate the solidification and microstructural evolution of LEU-10wt.%Mo. Electron backscatter diffraction (EBSD) and complementary electron probe microanalysis (EPMA) reveal significant microsegregation present in the metastable γ-phase after solidification. Homogenization is performed at 800 and 1000 °C for times ranging from 1 to 32 h to explore the time–temperature combinations that will reduce the extent of microsegregation, as regions of higher and lower Mo content may influence local mechanical properties and provide preferred regions for γ-phase decomposition. We show for the first time that EBSD can be used to qualitatively assess microstructural evolution in DU-10wt.%Mo after homogenization treatments. Complementary EPMA is used to quantitatively confirm this finding. Homogenization at 1000 °C for 2–4 h may the regions that contain 8 wt.% Mo or lower, whereas homogenization at 1000 °C for longer than 8 h effectively saturates Mo chemical homogeneity, but results in substantial grain growth. The appropriate homogenization time will depend upon additional microstructural considerations, such as grain growth and intended subsequent processing. Higher carbon LEU-10wt.%Mo generally contains more inclusions within the grains and at grain boundaries after solidification. The effect of these inclusions on microstructural evolution (e.g. grain growth) during homogenization and as potential γ-phase decomposition nucleation sites is unclear, but likely requires additional study.

  7. Microstructural evolution of a uranium-10 wt.% molybdenum alloy for nuclear reactor fuels

    Energy Technology Data Exchange (ETDEWEB)

    Clarke, A.J., E-mail: aclarke@lanl.gov; Clarke, K.D.; McCabe, R.J.; Necker, C.T.; Papin, P.A.; Field, R.D.; Kelly, A.M.; Tucker, T.J.; Forsyth, R.T.; Dickerson, P.O.; Foley, J.C.; Swenson, H.; Aikin, R.M.; Dombrowski, D.E.

    2015-10-15

    Low-enriched uranium-10 wt.% molybdenum (LEU-10wt.%Mo) is of interest for the fabrication of monolithic fuels to replace highly-enriched uranium (HEU) dispersion fuels in high performance research and test reactors around the world. In this work, depleted uranium-10 wt.%Mo (DU-10wt.%Mo) is used to simulate the solidification and microstructural evolution of LEU-10wt.%Mo. Electron backscatter diffraction (EBSD) and complementary electron probe microanalysis (EPMA) reveal significant microsegregation present in the metastable γ-phase after solidification. Homogenization is performed at 800 and 1000 °C for times ranging from 1 to 32 h to explore the time–temperature combinations that will reduce the extent of microsegregation, as regions of higher and lower Mo content may influence local mechanical properties and provide preferred regions for γ-phase decomposition. We show for the first time that EBSD can be used to qualitatively assess microstructural evolution in DU-10wt.%Mo after homogenization treatments. Complementary EPMA is used to quantitatively confirm this finding. Homogenization at 1000 °C for 2–4 h may the regions that contain 8 wt.% Mo or lower, whereas homogenization at 1000 °C for longer than 8 h effectively saturates Mo chemical homogeneity, but results in substantial grain growth. The appropriate homogenization time will depend upon additional microstructural considerations, such as grain growth and intended subsequent processing. Higher carbon LEU-10wt.%Mo generally contains more inclusions within the grains and at grain boundaries after solidification. The effect of these inclusions on microstructural evolution (e.g. grain growth) during homogenization and as potential γ-phase decomposition nucleation sites is unclear, but likely requires additional study.

  8. Fabrication of uranium alloy fuel slug for sodium-cooled fast reactor by injection casting

    International Nuclear Information System (INIS)

    Jong Hwan Kim; Hoon Song; Ki Hwan Kim; Chan Bock Lee

    2014-01-01

    Metal fuel slugs of U-Zr alloys for a sodium-cooled fast reactor (SFR) have been fabricated using an injection casting method. However, casting alloys containing volatile radioactive constituents such as Am can cause problems in a conventional injection casting method. Therefore, in this study, several injection-casting methods were applied to evaluate the volatility of the metal-fuel elements and control the transport of volatile elements. Mn was selected as a volatile surrogate alloy since it possesses a total vapor pressure equivalent to that of minor actinide-bearing fuels for SFRs. U-10 wt% Zr and U-10 wt% Zr-5 wt% Mn metal fuels were prepared, and the casting processes were evaluated. The casting soundness of the fuel slugs was characterized by gamma-ray radiography and immersion density measurements. Inductively coupled plasma atomic emission spectroscopy was used to determine the chemical composition of fuel slugs. Fuel losses after casting were also evaluated according to the casting conditions. (author)

  9. Some aspects of in-pile swelling of fissile materials, 1. part: non-alloyed α uranium

    International Nuclear Information System (INIS)

    Mikailoff, H.

    1964-01-01

    An examination has been carried out of non-alloyed uranium samples, having various structural states, cold-worked and recrystallized, as-cast and β-treated, and irradiated at temperatures of between 450 and 600 C and with burn-ups from 1300 to 5500 MW days/metric ton. These samples swelled because of precipitation of the fission gases the porosity thus produced has a morphology depending mainly on the type of deformation to which the metal has been subjected and which is due to in-pile growth. The most homogeneous distribution of pores, and thus that leading to the minimum swelling, is only observed in the material having a marked [010] texture in which the growth and perhaps the thermal cycling introduce little or no strain. For other materials the deformation /swelling association causes a more rapid destruction of the samples either by cracking when the deformation is due to twinning, or by pronounced swelling localized in the bands when deformation is due to slipping. Finally the fission-gas precipitation considerably facilitates, above 500 C, the germination and growth of the intergranular cracks which can then develop at low stresses. (author) [fr

  10. 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)

  11. Manhattan Project Technical Series The Chemistry of Uranium (I) Chapters 1-10

    International Nuclear Information System (INIS)

    Rabinowitch, E. I.; Katz, J. J.

    1946-01-01

    This constitutes Chapters 1 through 10. inclusive, of The Survey Volume on Uranium Chemistry prepared for the Manhattan Project Technical Series. Chapters are titled: Nuclear Properties of Uranium; Properties of the Uranium Atom; Uranium in Nature; Extraction of Uranium from Ores and Preparation of Uranium Metal; Physical Properties of Uranium Metal; Chemical Properties of Uranium Metal; Intermetallic Compounds and Alloy systems of Uranium; the Uranium-Hydrogen System; Uranium Borides, Carbides, and Silicides; Uranium Nitrides, Phosphides, Arsenides, and Antimonides.

  12. Fabrication of powder from ductile uranium alloys for use as nuclear dispersion

    International Nuclear Information System (INIS)

    Durazzo, M.; Leal Neto, R.M.; Rocha, C.J.; Urano de Carvalho, E.; Riella, H.G.

    2014-01-01

    This work forms part of the studies presently ongoing at IPEN investigating the feasibility of powdering ductile U-10wt%Mo alloy by hydriding-milling-de-hydriding of the gamma phase (HMD). Hydriding was conducted at room temperature in a Sievert apparatus following heat treatment activation. Hydrided pieces were fragile enough to be hand milled to the desired particle size range. Hydrogen was removed by heating the samples under high vacuum. X-ray diffraction analysis of the hydrided material showed an amorphous-like pattern that is completely reversed following de-hydriding. The hydrogen content of the hydrided samples corresponds to a trihydride, i.e. (U,Mo)H 3 . SEM analysis of HMD powder particles revealed equi-axial powder particles together with some plate-like particles. A hypothesis for the amorphous hydride phase formation is suggested. (authors)

  13. Thermal simulation of quenching uranium-0.75% titanium alloy in water

    International Nuclear Information System (INIS)

    Siman-Tov, M.; Llewellyn, G.H.; Childs, K.W.; Ludtka, G.M.; Aramayo, G.A.

    1985-01-01

    A computer model, The Quench Simulator, has been developed to simulate and predict in detail the behavior of U-0.75 Ti alloy when quenched at high temperature (about 850 0 C) in cold water. The code allows one to determine the time- and space-dependent distributions of temperature, residual stress, distortion, and microstructure that evolve during the quenching process. The nonlinear temperature- and microstructure-dependent properties, as well as the cooling rate-dependent heats of transformation, are incorporated into the model. The complex boiling heat transfer with its various regimes and other thermal boundary conditions are simulated. Experiments have been performed and incorporated into the model. Both sudden submersion and gradual controlled immersion can be applied. A parametric and sensitivity study has been performed demonstrating the importance of the thermal boundary conditions applied for achieving certain product characteristics. The thermal aspects of the model and its applications are discussed and demonstrated

  14. Biaxial Creep Specimen Fabrication

    International Nuclear Information System (INIS)

    JL Bump; RF Luther

    2006-01-01

    This report documents the results of the weld development and abbreviated weld qualification efforts performed by Pacific Northwest National Laboratory (PNNL) for refractory metal and superalloy biaxial creep specimens. Biaxial creep specimens were to be assembled, electron beam welded, laser-seal welded, and pressurized at PNNL for both in-pile (JOYO reactor, O-arai, Japan) and out-of-pile creep testing. The objective of this test campaign was to evaluate the creep behavior of primary cladding and structural alloys under consideration for the Prometheus space reactor. PNNL successfully developed electron beam weld parameters for six of these materials prior to the termination of the Naval Reactors program effort to deliver a space reactor for Project Prometheus. These materials were FS-85, ASTAR-811C, T-111, Alloy 617, Haynes 230, and Nirnonic PE16. Early termination of the NR space program precluded the development of laser welding parameters for post-pressurization seal weldments

  15. Biaxial Creep Specimen Fabrication

    Energy Technology Data Exchange (ETDEWEB)

    JL Bump; RF Luther

    2006-02-09

    This report documents the results of the weld development and abbreviated weld qualification efforts performed by Pacific Northwest National Laboratory (PNNL) for refractory metal and superalloy biaxial creep specimens. Biaxial creep specimens were to be assembled, electron beam welded, laser-seal welded, and pressurized at PNNL for both in-pile (JOYO reactor, O-arai, Japan) and out-of-pile creep testing. The objective of this test campaign was to evaluate the creep behavior of primary cladding and structural alloys under consideration for the Prometheus space reactor. PNNL successfully developed electron beam weld parameters for six of these materials prior to the termination of the Naval Reactors program effort to deliver a space reactor for Project Prometheus. These materials were FS-85, ASTAR-811C, T-111, Alloy 617, Haynes 230, and Nirnonic PE16. Early termination of the NR space program precluded the development of laser welding parameters for post-pressurization seal weldments.

  16. Extrusion of the uranium-0.75 weight percent titanium alloy

    International Nuclear Information System (INIS)

    Jackson, R.J.; Lundberg, M.R.; Boland, J.F.

    1975-01-01

    Procedures are described for extruding the U--0.75 wt percent Ti alloy in the high alpha region (600 to 640 0 C) , and in the upper gamma region (900 to 1000 0 C). The casting of sound extrusion billets has importance in the production of sound extrusions, and procedures are given for casting sound billets up to 1,100 kilograms . Also important in producing sound extrusions is the use of glass lubricants. Reduction ratios of greater than 50 to 1 were achieved on reasonably sized billets. Extrusion constants of 48,000 pounds per square inch (psi) [296 megapascals (MPa)] for alpha phase (630 0 C) and 8,000 psi (56 MPa) for gamma phase (950 0 C) were achieved. Gamma-phase extrusion has preference over alpha-phase extrusion in that larger billets can be used and temperature control is not as critical. However alpha-phase extrusion offers better surface finish, less die wear, and fewer oxidation problems. Billets up to 14 inches in diameter have been successfully gamma-extruded and plans exist for extruding billets up to 20 inches (508 millimetres) in diameter. (U.S.)

  17. Orientational relationships between phases in the γ→α transformations for uranium-molybdenum alloys

    International Nuclear Information System (INIS)

    Brun, G.

    1966-04-01

    A crystallographic study has been made of the γ → α + γ transformation in the alloy containing 3 per cent by weight of molybdenum using electronic micro-diffraction; it has been possible to establish the orientational relationships governing the germination of the α phase in the γ phase. One finds: (111)γ // (100) α, (112-bar)γ // (010) α, (11-bar 0)γ // (001)α. By choosing a monoclinic lattice containing the same number of atoms as the orthorhombic lattice for defining the γ mother phase, the change in structure has been explained by adding a homogeneous (112-bar)γ [111]γ shearing deformation to a heterogeneous deformation brought about by slipping of the atoms which are not situated at the nodes of this lattice. The identity of the orientation relationships γ/α and γ/α''b and the loss of coherence γ /α as a function of temperature or of time lead to the conclusion that, in the range studied, the γ → α transformation begins with a martensitic process and continues by germination and growth. (author) [fr

  18. Determination of trace impurities in uranium-transition metal alloy fuels by ICP-MS using extended common analyte internal standardization (ECAIS) technique

    International Nuclear Information System (INIS)

    Saha, Abhijit; Deb, S.B.; Nagar, B.K.; Saxena, M.K.

    2015-01-01

    An analytical methodology was developed for the determination of eight trace impurities viz, Al, B, Cd, Co, Cu, Mg, Mn and Ni in three different uranium-transition metal alloy fuels (U-Me; Me = Ti, Zr and Mo) employing inductively coupled plasma mass spectrometry (ICP-MS). The well known common analyte internal standardization (CAIS) chemometric technique was modified and then employed to minimize and account for the matrix effect on analyte intensity. Standard addition of analytes to the pure synthetic U-Me sample solutions and subsequently their ≥ 94% recovery by the ICP-MS measurement validates the proposed methodology. One real sample of each of these alloys was analyzed by the developed analytical methodology and the %RSD observed was in the range of 5-8%. The method detection limits were found to be within 4-10 μg L -1 . (author)

  19. Basic research on high-uranium density fuels for research and test reactors

    International Nuclear Information System (INIS)

    Ugajin, M.; Itoh, A.; Akabori, M.

    1992-01-01

    High-uranium density fuels, uranium silicides (U 3 Si 2 , U 3 Si) and U 6 Me-type uranium alloys (Me = Fe, Mn, Ni), were prepared and examined metallurgically as low-enriched uranium (LEU) fuels for research and test reactors. Miniature aluminum-dispersion plate-type fuel (miniplate) and aluminum-clad disk-type fuel specimens were fabricated and subjected to the neutron irradiation in JMTR (Japan Materials Testing Reactor). Fuel-aluminum compatibility tests were conducted to elucidate the extent of reaction and to identify reaction products. The relative stability of the fuels in an aluminum matrix was established at 350degC or above. Experiments were also performed to predict the chemical form of the solid fission-products in the uranium silicide (U 3 Si 2 ) simulating a high burnup anticipated for reactor service. (author)

  20. Interaction between uranium oxide alloyed with Al2O3·SiO2 and pyrocarbon coating during irradiation of micro fuel elements

    International Nuclear Information System (INIS)

    Chernikov, A.S.; Khromov, Y.F.; Svistunov, D.E.; Chuiko, E.E.

    1989-01-01

    The thermodynamics of the interaction between uranium oxide and carbon was previously studied in the presence of Al 2 O 3 ·SiO 2 , SiC, and UC 1.86 ; in this case, the quantity of the reacting substances does not have any effect on the attainment of the equilibrium state. Based on the obtained results, it is interesting to study the characteristic features of the interaction between the alloyed UO x cores (kernels) with the PyC-coating under the conditions involving irradiation of the micro fuel elements with thermal neutrons and the formation of solid fission products. The data concerning the characteristics of a micro fuel element (the weight of the core, its composition, etc.) are useful for carrying out a quantitative evaluation of the additives required for fixing the alkali-earth fission products by obtaining stable compounds of aluminosilicates with Ba, Sr, Rb, and Cs at different levels of depletion (burnup) of the oxide fuel. An analysis of the interaction processes in such a complex system as the irradiated alloyed uranium oxide fuel located in a micro fuel element is carried out by comparing the chemical potential of oxygen (RT ln P O 2 ) for the competing constituents of the system

  1. Influence of heat treatments for laser welded semi solid metal cast A356 alloy on the fracture mode of tensile specimens

    CSIR Research Space (South Africa)

    Kunene, G

    2008-09-01

    Full Text Available were then butt laser welded. It was found that the pre-weld as cast, T4 and post-weld T4 heat treated specimens fractured in the base metal. However, the pre-weld T6 heat treated specimens were found to have fractured in the heat affected zone (HAZ)...

  2. Thermal cycling behaviour and thermal stability of uranium-molybdenum alloys of low molybdenum content; Comportement au cyclage thermique et stabilite thermique des alliaces uranium-molybdene de faibles teneurs en molybdene

    Energy Technology Data Exchange (ETDEWEB)

    Decours, J; Fabrique, B; Peault, O [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1963-07-01

    We have studied the behaviour during thermal cycling of as-cast U-Mo alloys whose molybdenum content varies from 0.5 to 3 per cent; results are given concerning grain stability during extended heat treatments and the effect of treatments combining protracted heating with thermal cycling. The thermal cycling treatments were carried out at 550, 575, 600 and 625 deg C for 1000 cycles; the protracted heating experiments were done at 550, 575, 600 and 625 deg C for 2000 hours (4000 hrs at 625 deg C). The 0.5 per cent alloy resists much better to the thermal cycling than does the non-alloyed uranium. This resistance is, however, much lower than that of alloys containing over l per cent, even at 550 deg C it improves after a heat treatment for grain-refining. Alloys of over 1.1 per cent have a very good resistance to a cycling treatment even at 625 deg C, and this behaviour improves with increasing concentrations up to 3 per cent. An increase in the temperature up to the {gamma}-phase has few disadvantages provided that it is followed by rapid cooling (50 to 100 deg C/min). The {alpha} grain is fine, the {gamma}-phase is of the modular form, and the behaviour during a thermal cycling treatment is satisfactory. If this cooling is slow (15 deg /hr) the {alpha}-grain is coarse and cycling treatment behaviour is identical to that of the 0.5 per cent alloy. The protracted heat treatments showed that the {alpha}-grain exhibits satisfactory stability after 2000 hours at 575, 600 and 625 deg C, and after 4000 hours at 625 deg C. A heat cycling treatment carried out after these tests affects only very little the behaviour of these alloys during cycling. (authors) [French] Nous avons etudie le comportement au cyclage thermique des alliages U-Mo, brut de coulee, dont la teneur varie de 0,5 a 3 pour cent de molybdene, les resultats de stabilite du grain au cours de traitements thermiques de longue duree, ainsi que ceux des traitements combines de longue duree et de cyclage. Les

  3. A study of phase transformations processes in 0,5 to 4% mo uranium-molybdenum alloys; Etude des processus des transformations dans les alliages uranium-molybdene de teneur 0,5 a 4% en poids de molybdene

    Energy Technology Data Exchange (ETDEWEB)

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

    1959-06-15

    Isothermal and continuous cooling transformations process have been established on uranium-molybdenum alloys containing 0,5 to 4 w% Mo. Transformations process of the {beta} and {gamma} solid solutions are described. These processes depend upon molybdenum concentration. Out of the {beta} solid solution phase appears an eutectoid decomposition of {beta} to ({alpha} + {gamma}) or the formation of a martensitic phase {alpha}''. The {gamma} solid solution shows a decomposition of {gamma} to ({alpha} + {gamma}) or ({alpha} + {gamma}'), or a formation of martensitic phases a' or a'{sub b}. The U-Mo equilibrium diagram is discussed, particularly in low concentrations zones. Limits between domains ({alpha} + {gamma}) and ({beta} + {gamma}), ({beta} + {gamma}) and {gamma}, ({beta} + {gamma}) and {beta}, have been determined. (author) [French] Les processus des transformations isothermes, et au cours de refroidissements continus ont ete etablis sur les alliages uranium-molybdene de 0,5 a 4 % en poids de Mo. Ceci a permis de mettre en evidence les processus des transformations de solutions solides {beta} et {gamma}, differents suivant la teneur en molybdene de l'alliage. Dans le premier cas il y a decomposition eutectoide de {beta} en ({alpha} + {gamma}) ou formations d'une phase martensitique {alpha}''. Dans le second cas il y a decomposition de {gamma} soit en ({alpha} + {gamma}) soit en ({alpha} + {gamma}') suivant la temperature, ou bien formation des phases martensitiques {alpha}' ou {alpha}'{sub b}. Le diagramme d'equilibre, uranium-molybdene est sujet a de nombreuses controverses, en particulier dans la zone des faibles concentrations. Les limites entre les domaines ({alpha} + {gamma}) et ({beta} + {gamma}), ({beta} + {gamma}) et {gamma}, ({beta} + {gamma}) et {beta}, ont ete determinees. (auteur)

  4. Technique for preparation of transmission electron microscope specimens from wire samples of Al and Al-Al2O3 alloys

    DEFF Research Database (Denmark)

    Lindbo, Jørgen

    1966-01-01

    A technique for thinning 1 mm wire samples of aluminium and aluminium-alumina alloys for transmission electron microscopy is described. The essential feature of the technique, which involves spark machining and electropolishing in a polytetrafluoroethylene holder followed by chemical polishing...

  5. Effect of CT Specimen Thickness on the Mechanical Characteristics at the Crack Tip of Stress Corrosion Cracking in Ni-based Alloys

    Science.gov (United States)

    Yinghao, Cui; He, Xue; Lingyan, Zhao

    2017-12-01

    It’s important to obtain accurate stress corrosion crack(SCC) growth rate for quantitative life prediction of components in nuclear power plants. However, the engineering practice shows that the crack tip constraint effect has a great influence on the mechanical properties and crack growth rate of SCC at crack tip. To study the influence of the specimen thickness on the crack tip mechanical properties of SCC, the stress, strain and C integral at creep crack tip are analyzed under different specimens thickness. Results show that the cracked specimen is less likely to crack due to effect of crack tip constraint. When the thickness ratio B/W is larger than 0.1, the crack tip constraint is almost ineffective. Value of C integral is the largest when B/W is 0.25. Then specimen thickness has little effect on the value of C integral. The effect of specimen thickness on the value of C integral is less significant at higher thickness ratio.

  6. Development of a program in LABVIEW platform to controlling and monitoring a Sievert-type system for comminution of metallic uranium and its alloys

    International Nuclear Information System (INIS)

    Dutra, Aimore R.R.; Ferraz, Wilmar B.; Ferreira, Ricardo A.N.

    2011-01-01

    A comminution process by hydriding-dehydriding method was developed at CDTN-Centro de Desenvolvimento da Tecnologia Nuclear with the purpose of obtaining plate type nuclear fuel. This fuel requires the use of metallic uranium and its alloys in form of powders. This comminution process was performed based on a Sievert system. Initially this system was controlled and monitored by a computer program developed in Turbo Pascal language. In order to improve the control of the comminution process, a new program was developed in LabVIEW platform. This paper presents a description of this new program and the main aspects of the operation of the system. The more accurate monitoring and controlling of the various stages of the comminution process as well as greater flexibility in the choice of input data, real-time graphics, generation of reports and a reduction of time passivation were achieved. (author)

  7. Development of a program in LABVIEW platform to controlling and monitoring Sievert-type system for comminution of metallic uranium and its alloys

    International Nuclear Information System (INIS)

    Dutra, Aimore R.R.; Ferraz, Wilmar B.; Ferreira, Ricardo A.N.

    2011-01-01

    A comminution process by hydriding-de hydriding method was developed at CDTN-Centro de Desenvolvimento da Tecnologia Nuclear with the purpose of obtaining plate type nuclear fuel. This fuel requires the use of metallic uranium and its alloys in form of powders. This comminution process was performed based on a Sievert system. Initially this system was controlled and monitored by a computer program developed in Turbo Pascal language. In order to improve the control of the comminution process, a new program was developed in LabVIEW platform. This paper presents a description of this new program and the main aspects of the operation of the system. The more accurate monitoring and controlling of the various stages of the comminution process as well as greater flexibility in the choice of input data, real-time graphics, generation of reports and a reduction of time passivation were achieved. (author)

  8. Interaction of Al2O3xSiO2 alloyed uranium oxide with pyrocarbon coating of fuel particles under irradiation

    International Nuclear Information System (INIS)

    Chernikov, A.S.; Khromov, Yu.F.; Svistunov, D.E.; Chujko, E.E.

    1989-01-01

    Method of comparative data analysis for P O2 and P CO was used to consider interaction in fuel particle between pyrocarbon coating and fuel sample, alloyed with alumosilicate addition. Equations of interaction reactions for the case of hermetic and depressurized fuel particle are presented. Calculations of required xAl 2 O 3 XySiO 2 content, depending on oxide fuel burnup, were conducted. It was suggested to use silicon carbide for limitation of the upper level of CO pressure in fuel particle. Estimation of thermal stability of alumosilicates under conditions of uranium oxide burnup equals 1100 and 1500 deg C for Al/Si ratio in addition 1/1 and 4/1 respectively

  9. Development of a program in LABVIEW platform to controlling and monitoring a Sievert-type system for comminution of metallic uranium and its alloys

    Energy Technology Data Exchange (ETDEWEB)

    Dutra, Aimore R.R.; Ferraz, Wilmar B.; Ferreira, Ricardo A.N., E-mail: ferrazw@cdtn.b, E-mail: ranf@cdtn.b [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

    2011-07-01

    A comminution process by hydriding-dehydriding method was developed at CDTN-Centro de Desenvolvimento da Tecnologia Nuclear with the purpose of obtaining plate type nuclear fuel. This fuel requires the use of metallic uranium and its alloys in form of powders. This comminution process was performed based on a Sievert system. Initially this system was controlled and monitored by a computer program developed in Turbo Pascal language. In order to improve the control of the comminution process, a new program was developed in LabVIEW platform. This paper presents a description of this new program and the main aspects of the operation of the system. The more accurate monitoring and controlling of the various stages of the comminution process as well as greater flexibility in the choice of input data, real-time graphics, generation of reports and a reduction of time passivation were achieved. (author)

  10. Development of a program in LABVIEW platform to controlling and monitoring Sievert-type system for comminution of metallic uranium and its alloys

    Energy Technology Data Exchange (ETDEWEB)

    Dutra, Aimore R.R.; Ferraz, Wilmar B.; Ferreira, Ricardo A.N., E-mail: ferrazw@cdtn.b, E-mail: ranf@cdtn.b [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

    2011-07-01

    A comminution process by hydriding-de hydriding method was developed at CDTN-Centro de Desenvolvimento da Tecnologia Nuclear with the purpose of obtaining plate type nuclear fuel. This fuel requires the use of metallic uranium and its alloys in form of powders. This comminution process was performed based on a Sievert system. Initially this system was controlled and monitored by a computer program developed in Turbo Pascal language. In order to improve the control of the comminution process, a new program was developed in LabVIEW platform. This paper presents a description of this new program and the main aspects of the operation of the system. The more accurate monitoring and controlling of the various stages of the comminution process as well as greater flexibility in the choice of input data, real-time graphics, generation of reports and a reduction of time passivation were achieved. (author)

  11. Radiation damage of uranium; Radijaciono ostecenje urana

    Energy Technology Data Exchange (ETDEWEB)

    Lazarevic, Dj [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Yugoslavia)

    1966-11-15

    Study of radiation damage covered the following: Kinetics of electric resistance of uranium and uranium alloy with 1% of molybdenum dependent on the second phase and burnup rate; Study of gas precipitation and diffusion of bubbles by transmission electron microscopy; Numerical analysis of the influence of defects distribution and concentration on the rare gas precipitation in uranium; study of thermal sedimentation of uranium alloy with molybdenum; diffusion of rare gas in metal by gas chromatography method.

  12. Development of metal uranium fuel and testing of construction materials (I-VI); Part I

    International Nuclear Information System (INIS)

    Mihajlovic, A.

    1965-11-01

    This project includes the following tasks: Study of crystallisation of metal melt and beta-alpha transforms in uranium and uranium alloys; Study of the thermal treatment influence on phase transformations and texture in uranium alloys; Radiation damage of metal uranium; Project related to irradiation of metal uranium in the reactor; Development of fuel element for nuclear reactors

  13. Development of a high density fuel based on uranium-molybdenum alloys with high compatibility in high temperatures; Desenvolvimento de um combustivel de alta densidade a base das ligas uranio-molibdenio com alta compatibilidade em altas temperaturas

    Energy Technology Data Exchange (ETDEWEB)

    Oliveira, Fabio Branco Vaz de

    2008-07-01

    This work has as its objective the development of a high density and low enriched nuclear fuel based on the gamma-UMo alloys, for utilization where it is necessary satisfactory behavior in high temperatures, considering its utilization as dispersion. For its accomplishment, it was started from the analysis of the RERTR ('Reduced Enrichment for Research and Test Reactors') results and some theoretical works involving the fabrication of gamma-uranium metastable alloys. A ternary addition is proposed, supported by the properties of binary and ternary uranium alloys studied, having the objectives of the gamma stability enhancement and an ease to its powder fabrication. Alloys of uranium-molybdenum were prepared with 5 to 10% Mo addition, and 1 and 3% of ternary, over a gamma U7Mo binary base alloy. In all the steps of its preparation, the alloys were characterized with the traditional techniques, to the determination of its mechanical and structural properties. To provide a process for the alloys powder obtention, its behavior under hydrogen atmosphere were studied, in thermo analyser-thermo gravimeter equipment. Temperatures varied from the ambient up to 1000 deg C, and times from 15 minutes to 16 hours. The results validation were made in a semi-pilot scale, where 10 to 50 g of powders of some of the alloys studied were prepared, under static hydrogen atmosphere. Compatibility studies were conducted by the exposure of the alloys under oxygen and aluminum, to the verification of possible reactions by means of differential thermal analysis. The alloys were exposed to a constant heat up to 1000 deg C, and their performances were evaluated in terms of their reaction resistance. On the basis of the results, it was observed that ternary additions increases the temperatures of the reaction with aluminum and oxidation, in comparison with the gamma UMo binaries. A set of conditions to the hydration of the alloys were defined, more restrictive in terms of temperature

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

    International Nuclear Information System (INIS)

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

    2011-01-01

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

  15. Intergranular tellurium cracking of nickel-based alloys in molten Li, Be, Th, U/F salt mixture

    Science.gov (United States)

    Ignatiev, Victor; Surenkov, Alexander; Gnidoy, Ivan; Kulakov, Alexander; Uglov, Vadim; Vasiliev, Alexander; Presniakov, Mikhail

    2013-09-01

    In Russia, R&D on Molten Salt Reactor (MSR) are concentrated now on fast/intermediate spectrum concepts which were recognized as long term alternative to solid fueled fast reactors due to their attractive features: strong negative feedback coefficients, easy in-service inspection, and simplified fuel cycle. For high-temperature MSR corrosion of the metallic container alloy in primary circuit is the primary concern. Key problem receiving current attention include surface fissures in Ni-based alloys probably arising from fission product tellurium attack. This paper summarizes results of corrosion tests conducted recently to study effect of oxidation state in selected fuel salt on tellurium attack and to develop means of controlling tellurium cracking in the special Ni-based alloys recently developed for molten salt actinide recycler and tranforming (MOSART) system. Tellurium corrosion of Ni-based alloys was tested at temperatures up to 750 °C in stressed and unloaded conditions in molten LiF-BeF2 salt mixture fueled by about 20 mol% of ThF4 and 2 mol% of UF4 at different [U(IV)]/[U(III)] ratios: 0.7, 4, 20, 100 and 500. Following Ni-based alloys (in mass%): HN80М-VI (Mo—12, Cr—7.6, Nb—1.5), HN80МТY (Mo—13, Cr—6.8, Al—1.1, Ti—0.9), HN80МТW (Mo—9.4, Cr—7.0, Ti—1.7, W—5.5) and ЕМ-721 (W—25.2, Cr—5.7, Ti—0.17) were used for the study in the corrosion facility. If the redox state the fuel salt is characterized by uranium ratio [U(IV)]/[U(III)] uranium intermetallic compounds and alloys with nickel and molybdenum. This leads to spontaneous behavior of alloy formation processes on the specimens' surface and further diffusion of uranium deep into the metallic phase. As consequence of this films of intermetallic compounds and alloys of nickel, molybdenum, tungsten with uranium are formed on the alloys specimens' surface, and intergranular corrosion does not take place. In the fuel salt with [U(IV)]/[U(III)] = 4-20 the potentials of uranium

  16. Studies of plutonium-iron and uranium-plutonium-iron alloys; Etudes d'alliages plutonium-fer et d'alliages uranium-plutonium-fer

    Energy Technology Data Exchange (ETDEWEB)

    Avivi, Ehud [Commissariat a l' Energie Atomique, Fontenay-aux-Roses (France). Centre d' Etudes Nucleaires

    1964-01-15

    We study the plutonium-iron system, by means of dilatometry, X rays and metallography, especially in the domain between PuFe{sub 2} and Fe. We determine the solubilities of Fe in PuFe{sub 2} and of Pu in Fe. We show the presence of an hexagonal PuFe{sub 2} phase and we propose a modification in the Pu-Fe phase diagram. Some low iron concentration U-Pu-Fe alloys have also been investigated. We characterise the different phases. We confirm that adding some iron lowers the quantity of the zeta U-Pu phase. We emphasize some characteristics of the alloys having the global concentration (U, Pu){sub 6} Fe. (authors) [French] On etudie par dilatometrie, rayons X et micrographie le systeme plutonium-fer, principalement dans la region comprise entre PuFe{sub 2} et Fe, On determine les solubilites du fer dans PuFe{sub 2}, et de Pu dans Fe. On met en evidence une phase PuFe{sub 2} hexagonale et on propose une modification du diagramme d'equilibre Pu-Fe. Certains alliages U-Pu-Fe a faibles concentrations en fer sont egalement etudies. On caracterise les phases en presence. On confirme que l'addition de fer diminue rapidement la quantite de phase U-Pu zeta. Enfin on revele certaines caracteristiques des alliages de composition globale (U, Pu){sub 6} Fe. (auteurs)

  17. METHOD OF APPLYING NICKEL COATINGS ON URANIUM

    Science.gov (United States)

    Gray, A.G.

    1959-07-14

    A method is presented for protectively coating uranium which comprises etching the uranium in an aqueous etching solution containing chloride ions, electroplating a coating of nickel on the etched uranium and heating the nickel plated uranium by immersion thereof in a molten bath composed of a material selected from the group consisting of sodium chloride, potassium chloride, lithium chloride, and mixtures thereof, maintained at a temperature of between 700 and 800 deg C, for a time sufficient to alloy the nickel and uranium and form an integral protective coating of corrosion-resistant uranium-nickel alloy.

  18. Irradiation performance of uranium-molybdenum alloy dispersion fuels; Desempenho sob irradiacao de elementos combustiveis do tipo U-Mo

    Energy Technology Data Exchange (ETDEWEB)

    Almeida, Cirila Tacconi de

    2005-07-01

    The U-Mo-Al dispersion fuels of Material Test Reactors (MTR) are analyzed in terms of their irradiation performance. The irradiation performance aspects are associated to the neutronic and thermal hydraulics aspects to propose a new core configuration to the IEA-R1 reactor of IPEN-CNEN/SP using U-Mo-Al fuels. Core configurations using U-10Mo-Al fuels with uranium densities variable from 3 to 8 gU/cm{sup 3} were analyzed with the computational programs Citation and MTRCR-IEA R1. Core configurations for fuels with uranium densities variable from 3 to 5 gU/cm{sup 3} showed to be adequate to use in IEA-R1 reactor e should present a stable in reactor performance even at high burn-up. (author)

  19. Development and Validation of Capabilities to Measure Thermal Properties of Layered Monolithic U-Mo Alloy Plate-Type Fuel

    Science.gov (United States)

    Burkes, Douglas E.; Casella, Andrew M.; Buck, Edgar C.; Casella, Amanda J.; Edwards, Matthew K.; MacFarlan, Paul J.; Pool, Karl N.; Smith, Frances N.; Steen, Franciska H.

    2014-07-01

    The uranium-molybdenum (U-Mo) alloy in a monolithic form has been proposed as one fuel design capable of converting some of the world's highest power research reactors from the use of high enriched uranium to low enriched uranium. One aspect of the fuel development and qualification process is to demonstrate appropriate understanding of the thermal-conductivity behavior of the fuel system as a function of temperature and expected irradiation conditions. The purpose of this paper is to verify functionality of equipment installed in hot cells for eventual measurements on irradiated uranium-molybdenum (U-Mo) monolithic fuel specimens, refine procedures to operate the equipment, and validate models to extract the desired thermal properties. The results presented here demonstrate the adequacy of the equipment, procedures, and models that have been developed for this purpose based on measurements conducted on surrogate depleted uranium-molybdenum (DU-Mo) alloy samples containing a Zr diffusion barrier and clad in aluminum alloy 6061 (AA6061). The results are in excellent agreement with thermal property data reported in the literature for similar U-Mo alloys as a function of temperature.

  20. Vacuum fusion of uranium

    International Nuclear Information System (INIS)

    Stohr, J.A.

    1957-01-01

    After having outlined that vacuum fusion and moulding of uranium and of its alloys have some technical and economic benefits (vacuum operations avoid uranium oxidation and result in some purification; precision moulding avoids machining, chip production and chemical reprocessing of these chips; direct production of the desired shape is possible by precision moulding), this report presents the uranium fusion unit (its low pressure enclosure and pumping device, the crucible-mould assembly, and the MF supply device). The author describes the different steps of cast production, and briefly comments the obtained results

  1. Influence of specimen size and grain orientation to the life of a polycrystalline Ni-base alloy at LCF stress; Einfluss der Probengroesse und der Kornorientierung auf die Lebensdauer einer polykristallinen Ni-Basislegierung bei LCF-Beanspruchung

    Energy Technology Data Exchange (ETDEWEB)

    Seibel, Thomas

    2014-07-01

    In the present work the LCF (Low Cycle Fatigue) crack initiation life of the conventionally cast Ni-base alloy RENE 80 was analyzed as a function of specimen size and grain orientation. Five specimen geometries with distinctly different gauge sections were used: 3 geometries with cylindrical gauge section (G1-G3) and two notched geometries with a stress concentration factor of α{sub 1} = 1,62 (KG1) and α{sub 2} = 2,60 (KG2), resulting in a maximum difference of the damage relevant surface area up to a factor of approximately 72. Correction factors were determined by FEM calculations for all specimen geometries with highly reduced gauge sections where direct strain measurement was not possible. Additionally a uniform failure criterion with a relatively small crack size of 0,962 mm{sup 2} was defined. Totally, 116 isothermal LCF tests were carried out at the different specimen types at a temperature of 850 C in total strain control with a load ratio (minimum strain / maximum strain) of R{sub ε} = -1. The load cycles were applied with triangular waveform at a frequency of 0.1 Hz for high strain amplitudes and 1 Hz for low strain amplitudes, respectively. After the LCF-Tests the fracture surfaces of all samples were analyzed in more detail by SEM to identify the crack initiation mechanisms as well as the crack initiation sites. In this context it could be shown, that fatigue cracks were generally initiated at slip bands in surface grains. Accordingly, the grain orientations at the crack initiation sites were measured by electron back scatter diffraction (EBSD) and the maximum shear stresses in the respective principal slip system (111) <110> was calculated using the Schmid approach. For this, longitudinal sections were be prepared exactly at the crack initiation sites of samples loaded with low strain amplitudes where clearly defined single crack initiation sites were observed. Afterwards the maximum shear stress in the principal slip system at the crack initiation

  2. Some aspects of in-pile swelling of fissile materials, 1. part: non-alloyed {alpha} uranium; Quelques aspects du gonflement en pile des materiaux fissiles. 1. partie: uranium {alpha} non allie

    Energy Technology Data Exchange (ETDEWEB)

    Mikailoff, H [Commissariat a l' Energie Atomique, Fontenay-aux-Roses (France). Centre d' Etudes Nucleaires

    1964-07-01

    An examination has been carried out of non-alloyed uranium samples, having various structural states, cold-worked and recrystallized, as-cast and {beta}-treated, and irradiated at temperatures of between 450 and 600 C and with burn-ups from 1300 to 5500 MW days/metric ton. These samples swelled because of precipitation of the fission gases the porosity thus produced has a morphology depending mainly on the type of deformation to which the metal has been subjected and which is due to in-pile growth. The most homogeneous distribution of pores, and thus that leading to the minimum swelling, is only observed in the material having a marked [010] texture in which the growth and perhaps the thermal cycling introduce little or no strain. For other materials the deformation /swelling association causes a more rapid destruction of the samples either by cracking when the deformation is due to twinning, or by pronounced swelling localized in the bands when deformation is due to slipping. Finally the fission-gas precipitation considerably facilitates, above 500 C, the germination and growth of the intergranular cracks which can then develop at low stresses. (author) [French] On a examine des echantillons d'uranium non allie, de divers etats structuraux, marteles et recristallises, bruts de coulee et traites {beta}, irradies a des temperatures comprises entre 450 et 600 C, et a des taux de combustion allant de 1300 a 5500 MWj/t. Ces echantillons ont gonfle par suite de la precipitation de gaz de fission: la porosite ainsi fournie a une morphologie qui depend principalement des modes de deformation subie par le metal et due a la croissance en pile. La repartition la plus homogene des pores, donc celle qui donnera le gonflement minimum, est observee seulement dans le materiau a forte texture [010] dans lequel la croissance et eventuellement le cyclage thermique introduisent peu ou pas de contraintes. Dans les autres materiaux l'association deformation/gonflement rend plus rapide

  3. Uranium absorption study pile

    International Nuclear Information System (INIS)

    Raievski, V.; Sautiez, B.

    1959-01-01

    The report describes a pile designed to measure the absorption of fuel slugs. The pile is of graphite and comprises a central section composed of uranium rods in a regular lattice. RaBe sources and BF 3 counters are situated on either side of the center. A given uranium charge is compared with a specimen charge of about 560 kg, and the difference in absorption between the two noted. The sensitivity of the equipment will detect absorption variations of about a few ppm boron (10 -6 boron per gr. of uranium) or better. (author) [fr

  4. Fabrication and characterisation of uranium, molybdenum, chromium, niobium and aluminium; Dobijanje i karakterizacija legura uranijuma sa molibdenom, hromom, niobijumom i aluminijumom

    Energy Technology Data Exchange (ETDEWEB)

    Sofrenovic, R; Isailovic, M; Kotur, Z [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Yugoslavia)

    1965-11-15

    This paper describes fabrication of binary uranium alloys by melting and casting. The following alloys with nominal composition were obtained by melting in the vacuum furnace: uranium with niobium contents from 0.5%- 4.0% and uranium with molybdenum contents from 0.4% - 1.2%. Uranium alloys with chromium content from 0.4% - 1.2% and uranium alloy with 0.12% of aluminium were obtained by vacuum induction furnace (electric arc melting)

  5. Development of metal uranium fuel and testing of construction materials (I-VI); Part I; Razvoj metalnog goriva i ispitivanje konstrukcionih materijala (I-VI deo); I deo

    Energy Technology Data Exchange (ETDEWEB)

    Mihajlovic, A [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Yugoslavia)

    1965-11-15

    This project includes the following tasks: Study of crystallisation of metal melt and beta-alpha transforms in uranium and uranium alloys; Study of the thermal treatment influence on phase transformations and texture in uranium alloys; Radiation damage of metal uranium; Project related to irradiation of metal uranium in the reactor; Development of fuel element for nuclear reactors.

  6. Analysis of intergranular fission-gas bubble-size distributions in irradiated uranium-molybdenum alloy fuel

    Energy Technology Data Exchange (ETDEWEB)

    Rest, J. [Argonne National Laboratory, 9700 S. Cass Avenue, Argonne, IL 60439 (United States)], E-mail: jrest@anl.gov; Hofman, G.L.; Kim, Yeon Soo [Argonne National Laboratory, 9700 S. Cass Avenue, Argonne, IL 60439 (United States)

    2009-04-15

    An analytical model for the nucleation and growth of intra and intergranular fission-gas bubbles is used to characterize fission-gas bubble development in low-enriched U-Mo alloy fuel irradiated in the advanced test reactor in Idaho as part of the Reduced Enrichment for Research and Test Reactor (RERTR) program. Fuel burnup was limited to less than {approx}7.8 at.% U in order to capture the fuel-swelling stage prior to irradiation-induced recrystallization. The model couples the calculation of the time evolution of the average intergranular bubble radius and number density to the calculation of the intergranular bubble-size distribution based on differential growth rate and sputtering coalescence processes. Recent results on TEM analysis of intragranular bubbles in U-Mo were used to set the irradiation-induced diffusivity and re-solution rate in the bubble-swelling model. Using these values, good agreement was obtained for intergranular bubble distribution compared against measured post-irradiation examination (PIE) data using grain-boundary diffusion enhancement factors of 15-125, depending on the Mo concentration. This range of enhancement factors is consistent with values obtained in the literature.

  7. Analysis of intergranular fission-gas bubble-size distributions in irradiated uranium-molybdenum alloy fuel

    Science.gov (United States)

    Rest, J.; Hofman, G. L.; Kim, Yeon Soo

    2009-04-01

    An analytical model for the nucleation and growth of intra and intergranular fission-gas bubbles is used to characterize fission-gas bubble development in low-enriched U-Mo alloy fuel irradiated in the advanced test reactor in Idaho as part of the Reduced Enrichment for Research and Test Reactor (RERTR) program. Fuel burnup was limited to less than ˜7.8 at.% U in order to capture the fuel-swelling stage prior to irradiation-induced recrystallization. The model couples the calculation of the time evolution of the average intergranular bubble radius and number density to the calculation of the intergranular bubble-size distribution based on differential growth rate and sputtering coalescence processes. Recent results on TEM analysis of intragranular bubbles in U-Mo were used to set the irradiation-induced diffusivity and re-solution rate in the bubble-swelling model. Using these values, good agreement was obtained for intergranular bubble distribution compared against measured post-irradiation examination (PIE) data using grain-boundary diffusion enhancement factors of 15-125, depending on the Mo concentration. This range of enhancement factors is consistent with values obtained in the literature.

  8. Urine culture - catheterized specimen

    Science.gov (United States)

    Culture - urine - catheterized specimen; Urine culture - catheterization; Catheterized urine specimen culture ... urinary tract infections may be found in the culture. This is called a contaminant. You may not ...

  9. Miniature tensile test specimens for fusion reactor irradiation studies

    International Nuclear Information System (INIS)

    Klueh, R.L.

    1985-01-01

    Three miniature sheet-type tensile specimens and a miniature rod-type specimen are being used to determine irradiated tensile properties for alloy development for fusion reactors. The tensile properties of type 316 stainless steel were determined with these different specimens, and the results were compared. Reasonably good agreement was observed. However, there were differences that led to recommendations on which specimens are preferred. 4 references, 9 figures, 6 tables

  10. Elastic-plastic analysis of the SS-3 tensile specimen

    International Nuclear Information System (INIS)

    Majumdar, S.

    1998-01-01

    Tensile tests of most irradiated specimens of vanadium alloys are conducted using the miniature SS-3 specimen which is not ASTM approved. Detailed elastic-plastic finite element analysis of the specimen was conducted to show that, as long as the ultimate to yield strength ratio is less than or equal to 1.25 (which is satisfied by many irradiated materials), the stress-plastic strain curve obtained by using such a specimen is representative of the true material behavior

  11. Specimen size effects in Charpy impact testing

    International Nuclear Information System (INIS)

    Alexander, D.J.; Klueh, R.L.

    1989-01-01

    Full-size , half-size, and third-size specimens from several different steels have been tested as part of an ongoing alloy development program. The smaller specimens permit more specimens to be made from small trail heats and are much more efficient for irradiation experiments. The results of several comparisons between the different specimen sizes have shown that the smaller specimens show qualitatively similar behavior to large specimens, although the upper-shelf energy level and ductile-to-ductile transition temperature are reduced. The upper-shelf energy levels from different specimen sizes can be compared by using a simple volume normalization method. The effect of specimen size and geometry on the ductile-to-ductile transition temperature is more difficult to predict, although the available data suggest a simple shift in the transition temperature due to specimen size changes.The relatively shallower notch used in smaller specimens alters the deformation pattern, and permits yielding to spread back to the notched surface as well as through to the back. This reduces the constraint and the peak stresses, and thus the initiation of cleavage is more difficult. A better understanding of the stress and strain distributions is needed. 19 refs., 3 figs., 3 tabs

  12. Microstructural characterization of an irradiated RERTR-6 U-7Mo/AA4043 alloy dispersion fuel plate specimen blister-tested to a final temperature of 500 °C

    Science.gov (United States)

    Keiser, Dennis D.; Jue, Jan-Fong; Gan, Jian; Miller, Brandon D.; Robinson, Adam B.; Madden, James W.; Ross Finlay, M.; Moore, Glenn; Medvedev, Pavel; Meyer, Mitch

    2017-05-01

    The Material Management and Minimization (M3) Reactor Conversion Program, in the past called the Reduced Enrichment for Research and Test Reactor (RERTR) Program, is developing low-enriched uranium (LEU) fuels for application in research and test reactors. U-Mo alloy dispersion fuel is one type being developed. Blister testing has been performed on different fuel plate samples to determine the margin to failure for fuel plates irradiated to different fission densities. Microstructural characterization was performed using scanning electron microscopy and transmission electron microscopy on a sample taken from a U-7Mo/AA4043 matrix dispersion fuel plate irradiated in the RERTR-6 experiment that was blister-tested up to a final temperature of 500 °C. The results indicated that two types of grain/cell boundaries were observed in the U-7Mo fuel particles, one with a relatively low Mo content and fission gas bubbles and a second type enriched in Si, due to interdiffusion from the Si-containing matrix, with little evidence of fission gas bubbles. With respect to the behavior of the major fission gas Xe, a significant amount of the Xe was still observed within the U-7Mo fuel particle, along with microns into the AA4043 matrix. For the fuel/matrix interaction layers that form during fabrication and then grow during irradiation, they change from the as-irradiated amorphous structure to one that is crystalline after blister testing. In the AA4043 matrix, the original Si-rich precipitates, which are typically observed in as-irradiated U-Mo dispersion fuel, get consumed due to interdiffusion with the U-7Mo fuel particles during the blister test. Finally, the fission gas bubbles that were originally around 3 nm in diameter and resided on a fission gas superlattice (FGS) in the intragranular regions of as-irradiated U-7Mo fuel grew in size (up to ∼20 nm diameter) during blister testing and, in many areas, are no longer organized as a superlattice.

  13. Microstructural characterization of an irradiated RERTR-6 U-7Mo/AA4043 alloy dispersion fuel plate specimen blister-tested to a final temperature of 500 °C

    Energy Technology Data Exchange (ETDEWEB)

    Keiser, Dennis D., E-mail: dennis.keiser@inl.gov [Nuclear Fuels and Materials Division, Idaho National Laboratory, P. O. Box 1625, Idaho Falls, ID 83415-6146 (United States); Jue, Jan-Fong; Gan, Jian; Miller, Brandon D.; Robinson, Adam B.; Madden, James W. [Nuclear Fuels and Materials Division, Idaho National Laboratory, P. O. Box 1625, Idaho Falls, ID 83415-6146 (United States); Ross Finlay, M. [Australian Nuclear Science and Technology Organization, PMB 1, Menai, NSW 2234 (Australia); Moore, Glenn; Medvedev, Pavel; Meyer, Mitch [Nuclear Fuels and Materials Division, Idaho National Laboratory, P. O. Box 1625, Idaho Falls, ID 83415-6146 (United States)

    2017-05-15

    The Material Management and Minimization (M3) Reactor Conversion Program, in the past called the Reduced Enrichment for Research and Test Reactor (RERTR) Program, is developing low-enriched uranium (LEU) fuels for application in research and test reactors. U–Mo alloy dispersion fuel is one type being developed. Blister testing has been performed on different fuel plate samples to determine the margin to failure for fuel plates irradiated to different fission densities. Microstructural characterization was performed using scanning electron microscopy and transmission electron microscopy on a sample taken from a U-7Mo/AA4043 matrix dispersion fuel plate irradiated in the RERTR-6 experiment that was blister-tested up to a final temperature of 500 °C. The results indicated that two types of grain/cell boundaries were observed in the U-7Mo fuel particles, one with a relatively low Mo content and fission gas bubbles and a second type enriched in Si, due to interdiffusion from the Si-containing matrix, with little evidence of fission gas bubbles. With respect to the behavior of the major fission gas Xe, a significant amount of the Xe was still observed within the U-7Mo fuel particle, along with microns into the AA4043 matrix. For the fuel/matrix interaction layers that form during fabrication and then grow during irradiation, they change from the as-irradiated amorphous structure to one that is crystalline after blister testing. In the AA4043 matrix, the original Si-rich precipitates, which are typically observed in as-irradiated U-Mo dispersion fuel, get consumed due to interdiffusion with the U-7Mo fuel particles during the blister test. Finally, the fission gas bubbles that were originally around 3 nm in diameter and resided on a fission gas superlattice (FGS) in the intragranular regions of as-irradiated U-7Mo fuel grew in size (up to ∼20 nm diameter) during blister testing and, in many areas, are no longer organized as a superlattice.

  14. Ultrasonic measurement of elastic moduli of 17-4 pH stainless steel and uranium -2 molybdenum from -400C to 8000C

    International Nuclear Information System (INIS)

    Gieske, J.H.

    1980-10-01

    Young's Modulus, shear modulus, and Poisson's ratio for 17-4 pH stainless steel and uranium -2 molybdenum are calculated from ultrasonic longitudinal and shear velocities determined from -40 0 C to 800 0 C. The ultrasonic velocities were determined at elevated temperatures using a through-transmission buffer rod arrangement. An indium-gallium slurry bond was used as an ultrasonic couplant between Cupernickel 10 alloy buffer rods and the specimen. Microstructural changes and phase transitions in the specimens are evident from the temperature dependence of the ultrasonic data. 10 figures, 3 tables

  15. Depleted uranium

    International Nuclear Information System (INIS)

    Huffer, E.; Nifenecker, H.

    2001-02-01

    This document deals with the physical, chemical and radiological properties of the depleted uranium. What is the depleted uranium? Why do the military use depleted uranium and what are the risk for the health? (A.L.B.)

  16. Small-scale Specimen Testing of Monolithic U-Mo Fuel Foils

    Energy Technology Data Exchange (ETDEWEB)

    Ramprashad Prabhakaran; Douglas E. Burkes; James I. Cole; Indrajit Charit; Daniel M. Wachs

    2008-10-01

    The objective of this investigation is to develop a shear punch testing (SPT) procedure and standardize it to evaluate the mechanical properties of irradiated fuels in a hot-cell so that the tensile behavior can be predicted using small volumes of material and at greatly reduced irradiation costs. This is highly important in the development of low-enriched uranium fuels for nuclear research and test reactors. The load-displacement data obtained using SPT can be interpreted in terms of and correlated with uniaxial mechanical properties. In order to establish a correlation between SPT and tensile data, sub-size tensile and microhardness testing were performed on U-Mo alloys. In addition, efforts are ongoing to understand the effect of test parameters (such as specimen thickness, surface finish, punch-die clearance, crosshead velocity and carbon content) on the measured mechanical properties, in order to rationalize the technique, prior to employing it on a material of unknown strength.

  17. Small-scale Specimen Testing of Monolithic U-Mo Fuel Foils

    International Nuclear Information System (INIS)

    Ramprashad Prabhakaran; Douglas E. Burkes; James I. Cole; Indrajit Charit; Daniel M. Wachs

    2008-01-01

    The objective of this investigation is to develop a shear punch testing (SPT) procedure and standardize it to evaluate the mechanical properties of irradiated fuels in a hot-cell so that the tensile behavior can be predicted using small volumes of material and at greatly reduced irradiation costs. This is highly important in the development of low-enriched uranium fuels for nuclear research and test reactors. The load-displacement data obtained using SPT can be interpreted in terms of and correlated with uniaxial mechanical properties. In order to establish a correlation between SPT and tensile data, sub-size tensile and microhardness testing were performed on U-Mo alloys. In addition, efforts are ongoing to understand the effect of test parameters (such as specimen thickness, surface finish, punch-die clearance, crosshead velocity and carbon content) on the measured mechanical properties, in order to rationalize the technique, prior to employing it on a material of unknown strength

  18. Determination of uranium traces in nuclear cans of nuclear reactors

    International Nuclear Information System (INIS)

    Acosta L, E.; Benavides M, A.M.; Sanchez P, L.

    1996-01-01

    To quantify the uranium content as impurity can be found in zirconium alloys and zircaloy, utilized to construct the sheaths containing fuels of the reactors of nuclear plants. The determination by fluorescence spectroscopy was employed as quality control measurement, at once the corrosion resistance, diminish with the increase of the uranium content in the alloys. (Author)

  19. Studies on yttrium oxide coatings for corrosion protection against molten uranium

    International Nuclear Information System (INIS)

    Chakravarthy, Y.; Bhandari, Subhankar; Pragatheeswaran; Thiyagarajan, T.K.; Ananthapadmanabhan, P.V.; Das, A.K.; Kumar, Jay; Kutty, T.R.G.

    2012-01-01

    Yttrium oxide is resistant to corrosion by molten uranium and its alloys. Yttrium oxide is recommended as a protective oxide layer on graphite and metal components used for melting and processing uranium and its alloys. This paper presents studies on the efficacy of plasma sprayed yttrium oxide coatings for barrier applications against molten uranium

  20. PROCESSING OF URANIUM-METAL-CONTAINING FUEL ELEMENTS

    Science.gov (United States)

    Moore, R.H.

    1962-10-01

    A process is given for recovering uranium from neutronbombarded uranium- aluminum alloys. The alloy is dissolved in an aluminum halide--alkali metal halide mixture in which the halide is a mixture of chloride and bromide, the aluminum halide is present in about stoichiometric quantity as to uranium and fission products and the alkali metal halide in a predominant quantity; the uranium- and electropositive fission-products-containing salt phase is separated from the electronegative-containing metal phase; more aluminum halide is added to the salt phase to obtain equimolarity as to the alkali metal halide; adding an excess of aluminum metal whereby uranium metal is formed and alloyed with the excess aluminum; and separating the uranium-aluminum alloy from the fission- productscontaining salt phase. (AEC)

  1. ELECTROCHEMICAL STUDIES OF URANIUM METAL CORROSION MECHANISM AND KINETICS IN WATER

    International Nuclear Information System (INIS)

    Boudanova, Natalya; Maslennikov, Alexander; Peretroukhine, Vladimir F.; Delegard, Calvin H.

    2006-01-01

    During long-term underwater storage of low burn-up uranium metal fuel, a corrosion product sludge forms containing uranium metal grains, uranium dioxide, uranates and, in some cases, uranium peroxide. Literature data on the corrosion of non-irradiated uranium metal and its alloys do not allow unequivocal prediction of the paragenesis of irradiated uranium in water. The goal of the present work conducted under the program 'CORROSION OF IRRADIATED URANIUM ALLOYS FUEL IN WATER' is to study the corrosion of uranium and uranium alloys and the paragenesis of the corrosion products during long-term underwater storage of uranium alloy fuel irradiated at the Hanford Site. The elucidation of the physico-chemical nature of the corrosion of irradiated uranium alloys in comparison with non-irradiated uranium metal and its alloys is one of the most important aspects of this work. Electrochemical methods are being used to study uranium metal corrosion mechanism and kinetics. The present part of work aims to examine and revise, where appropriate, the understanding of uranium metal corrosion mechanism and kinetics in water

  2. Uranium conversion

    International Nuclear Information System (INIS)

    Oliver, Lena; Peterson, Jenny; Wilhelmsen, Katarina

    2006-03-01

    FOI, has performed a study on uranium conversion processes that are of importance in the production of different uranium compounds in the nuclear industry. The same conversion processes are of interest both when production of nuclear fuel and production of fissile material for nuclear weapons are considered. Countries that have nuclear weapons ambitions, with the intention to produce highly enriched uranium for weapons purposes, need some degree of uranium conversion capability depending on the uranium feed material available. This report describes the processes that are needed from uranium mining and milling to the different conversion processes for converting uranium ore concentrate to uranium hexafluoride. Uranium hexafluoride is the uranium compound used in most enrichment facilities. The processes needed to produce uranium dioxide for use in nuclear fuel and the processes needed to convert different uranium compounds to uranium metal - the form of uranium that is used in a nuclear weapon - are also presented. The production of uranium ore concentrate from uranium ore is included since uranium ore concentrate is the feed material required for a uranium conversion facility. Both the chemistry and principles or the different uranium conversion processes and the equipment needed in the processes are described. Since most of the equipment that is used in a uranium conversion facility is similar to that used in conventional chemical industry, it is difficult to determine if certain equipment is considered for uranium conversion or not. However, the chemical conversion processes where UF 6 and UF 4 are present require equipment that is made of corrosion resistant material

  3. Uranium exploration

    International Nuclear Information System (INIS)

    De Voto, R.H.

    1984-01-01

    This paper is a review of the methodology and technology that are currently being used in varying degrees in uranium exploration activities worldwide. Since uranium is ubiquitous and occurs in trace amounts (0.2 to 5 ppm) in virtually all rocks of the crust of the earth, exploration for uranium is essentially the search of geologic environments in which geologic processes have produced unusual concentrations of uranium. Since the level of concentration of uranium of economic interest is dependent on the present and future price of uranium, it is appropriate here to review briefly the economic realities of uranium-fueled power generation. (author)

  4. Uranium alloy forming process research

    International Nuclear Information System (INIS)

    Chow, T.S.; Biesiada, T.A.; Sunwoo, A.; Long, J.; Anklam, T.; Kang, S.W.

    1997-01-01

    The study of modern U-6Nb processes is motivated by the needs to reduce fabrication costs and to improve efficiency in material usage. We have studied two potential options: physical vapor deposition (PVD) for manufacturing near-net-shape U-6Nb, and kinetic-energy metallization (KEM) as a supplemental process for refurbishing recycled parts. In FY 1996, we completed two series of PVD runs and heat treatment analyses, the characterization of the microstructure and mechanical properties, a comparison of the results to data for wrought-processed material, and experimental demonstration of the KEM feasibility process with a wide range of variables (particle materials and sizes, gases and gas pressures, and substrate materials), and computer modeling calculations

  5. 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.

  6. Laparoscopic specimen retrieval bags.

    Science.gov (United States)

    Smorgick, Noam

    2014-10-01

    Specimen retrieval bags have long been used in laparoscopic gynecologic surgery for contained removal of adnexal cysts and masses. More recently, the concerns regarding spread of malignant cells during mechanical morcellation of myoma have led to an additional use of specimen retrieval bags for contained "in-bag" morcellation. This review will discuss the indications for use retrieval bags in gynecologic endoscopy, and describe the different specimen bags available to date.

  7. Evaluation of refractory-metal-clad uranium nitride and uranium dioxide fuel pins after irradiation for times up to 10 450 hours at 990 C

    Science.gov (United States)

    Bowles, K. J.; Gluyas, R. E.

    1975-01-01

    The effects of some materials variables on the irradiation performance of fuel pins for a lithium-cooled space power reactor design concept were examined. The variables studied were UN fuel density, fuel composition, and cladding alloy. All pins were irradiated at about 990 C in a thermal neutron environment to the design fuel burnup. An 85-percent dense UN fuel gave the best overall results in meeting the operational goals. The T-111 cladding on all specimens was embrittled, possibly by hydrogen in the case of the UN fuel and by uranium and oxygen in the case of the UO2 fuel. Tests with Cb-1Zr cladding indicate potential use of this cladding material. The UO2 fueled specimens met the operational goals of less than 1 percent cladding strain, but other factors make UO2 less attractive than low-density UN for the contemplated space power reactor use.

  8. Thermal conductivity of uranium: effects of purity and microstructure

    International Nuclear Information System (INIS)

    Sandenaw, T.A.

    1975-10-01

    Thermal conductivity curves for polycrystalline uranium are presented for the temperature range below 373 0 K. The curves are for specimens prepared by different fabrication procedures from material of known purity and hardness. Included is a curve for U/2wt percent Mo alloy. Different mechanisms appear to be influencing the thermal conductivity behavior of uranium in well-defined temperature regions: below 37 to 43 0 K, approximately 40 to approximately 80 0 K, 80 to approximately 280 0 K, and from 280 0 K to the α → β transformation temperature. Mechanisms responsible for results in one temperature region continue to exert a strong influence in the next higher temperature region. Impurities and initial microstructure seem to influence results at any starting temperature. Evidence is presented for the possibility of imperfection ordering in uranium between approximately 40 and approximately 280 0 K. It is postulated that the type of ordering is capable with a martensite-like behavior and that all physical property results depend on the extent of a modification of the α-phase on cooling below approximately 280 0 K

  9. Evaluation of the electrochemical behavior of U{sub 2.5}Zr{sub 7.5}Nb and U{sub 3}Zr{sub 9}Nb uranium alloys in relation to the pH and the solution aeration

    Energy Technology Data Exchange (ETDEWEB)

    Mansur, Fabio Abud; Santos, Ana Maria Matildes dos; Ferraz, Wilmar Barbosa; Figueiredo, Celia de Araujo, E-mail: ferraz@cdtn.b [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

    2011-07-01

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

  10. Miniaturized fatigue crack growth specimen technology and results

    International Nuclear Information System (INIS)

    Puigh, R.J.; Bauer, R.E.; Ermi, A.M.; Chin, B.A.

    1981-01-01

    The miniature fatigue crack propagation technology has been extended to in-cell fabrication of irradiated specimens. Baseline testing of selected titanium alloys has been performed at 25 0 C in air. At relatively small values for the stress intensity factor, ΔK, the crack growth rates for all titanium alloys investigated are within a factor of three. The crack growth rates for these titanium alloys are a factor of three greater than the crack growth rates of either 316SS (20% CW) or HT-9. Each of the titanium alloys has observable crack propagation for stress intensity factors as small as 4.2 MPa√m

  11. Georeferencing Animal Specimen Datasets

    NARCIS (Netherlands)

    van Erp, M.G.J.; Hensel, R.; Ceolin, D.; van der Meij, M.

    2014-01-01

    For biodiversity research, the field of study that is concerned with the richness of species of our planet, it is of the utmost importance that the location of an animal specimen find is known with high precision. Due to specimens often having been collected over the course of many years, their

  12. Closeout of JOYO-1 Specimen Fabrication Efforts

    International Nuclear Information System (INIS)

    ME Petrichek; JL Bump; RF Luther

    2005-01-01

    Fabrication was well under way for the JOYO biaxial creep and tensile specimens when the NR Space program was canceled. Tubes of FS-85, ASTAR-811C, and T-111 for biaxial creep specimens had been drawn at True Tube (Paso Robles, CA), while tubes of Mo-47.5 Re were being drawn at Rhenium Alloys (Cleveland, OH). The Mo-47.5 Re tubes are now approximately 95% complete. Their fabrication and the quantities produced will be documented at a later date. End cap material for FS-85, ASTAR-811C, and T-111 had been swaged at Pittsburgh Materials Technology, Inc. (PMTI) (Large, PA) and machined at Vangura (Clairton, PA). Cutting of tubes, pickling, annealing, and laser engraving were in process at PMTI. Several biaxial creep specimen sets of FS-85, ASTAR-811C, and T-111 had already been sent to Pacific Northwest National Laboratory (PNNL) for weld development. In addition, tensile specimens of FS-85, ASTAR-811C, T-111, and Mo-47.5 Re had been machined at Kin-Tech (North Huntington, PA). Actual machining of the other specimen types had not been initiated. Flowcharts 1-3 detail the major processing steps each piece of material has experienced. A more detailed description of processing will be provided in a separate document [B-MT(SRME)-51]. Table 1 lists the in-process materials and finished specimens. Also included are current metallurgical condition of these materials and specimens. The available chemical analyses for these alloys at various points in the process are provided in Table 2

  13. High-temperature thermal conductivity of uranium chromite and uranium niobate

    International Nuclear Information System (INIS)

    Fedoseev, D.V.; Varshavskaya, I.G.; Lavrent'ev, A.V.; Oziraner, S.N.; Kuznetsova, D.G.

    1979-01-01

    The technique of determining thermal conductivity coefficient of uranium niobate and uranium chromite on heating with laser radiation is described. Determined is the coefficient of free-convective heat transfer (with provision for a conduction component) by means of a standard specimen. The thermal conductivity coefficients of uranium chromite and niobate were measured in the 1300-1700 K temperature range. The results are presented in a diagram form. It has been calculated, that the thermal conductivity coefficient for uranium niobate specimens is greater in comparison with uranium chromite specimens. The thermal conductivity coefficients of the materials mentioned depend on temperature very slightly. Thermal conductivity of the materials considerably depends on their porosity. The specimens under investigation were fabricated by the pressing method and had the following porosity: uranium chromite - 30 %, uranium niobate - 10 %. Calculation results show, that thermal conductivity of dense uranium chromite is higher than thermal conductivity of dense uranium niobate. The experimental error equals approximately 20 %, that is mainly due to the error of measuring the temperature equal to +-25 deg, with a micropyrometer

  14. Uranium determination in different compositions

    International Nuclear Information System (INIS)

    Bulyanitsa, L.S.; Ivanova, K.S.; Ryzhinskij, M.V.; Alekseeva, N.A.; Solntseva, L.F.; Shereshevskaya, I.I.

    1978-01-01

    For clarifying the suitability of two different methods of analysis for determining uranium without its previous purification, the analysis of uranium carbides (UC, UC 2 , UC - ZrC) and alloys (U - Al, U - Zr - Nb, U- Ti) has been carried out. Dissolution of the compositions examined was carried out either after previous calcining (UC, UC 2 ) or fusion with KHSO 4 (UC - ZrC), or in phosphoric acid (alloys). The first method, a variant of potentiometric titration, has been modified for small amounts of uranium. Titration was carried out on a semiautomatic titrating unit. The uranium amount per titration is about 4 to 5 mg. The second method of analysis is the coulombmetric titration at a constant current intensity. The quantity of uranium per titration was equal to 1 - 3 mg. The statistical processing of the results obtained was carried out by a dispersion analysis that allowed to reveal the influence of separate factors, such as method of analysis, type of composition, the non-uniformity of a sample, the enumerated factors influencing the dispersion of the analysis results. It has been shown that the both methods are equally suitable for analysis of the uranium compounds examined

  15. Some recent innovations in small specimen testing

    International Nuclear Information System (INIS)

    Odette, G.R.; He, M.; Gragg, D.; Klingensmith, D.; Lucas, G.E.

    2002-01-01

    New innovative small specimen test techniques are described. Finite element simulations show that combinations of cone indentation pile-up geometry and load-penetration depth relations can be used to determine both the yield stress and strain-hardening behavior of a material. Techniques for pre-cracking and testing sub-miniaturized fracture toughness bend bars, with dimensions of 1.65x1.65x9 mm 3 , or less, are described. The corresponding toughness-temperature curves have a very steep transition slope, primarily due to rapid loss of constraint, which has advantages in some experiments to characterize the effects of specified irradiation variables. As one example of using composite specimens, an approach to evaluating helium effects is proposed, involving diffusion bonding small wires of a 54 Fe-based ferritic-martensitic alloy to a surrounding fracture specimen composed of an elemental Fe-based alloy. Finally, we briefly outline some potential approaches to multipurpose specimens and test automation

  16. Recent advances in study of uranium surface chemistry in China

    Energy Technology Data Exchange (ETDEWEB)

    Luo, Lizhu; Lai, Xinchun [Science and Technology on Surface Physics and Chemistry Laboratory, Sichuan (China); Wang, Xiaolin [China Academy of Engineering Physics, Sichuan (China)

    2014-04-01

    Uranium is very important in nuclear energy industry; however, uranium and its alloys corrode seriously in various atmospheres because of their chemical reactivities. In China, continuous investigations focused on surface chemistry have been carried out for a thorough understanding of uranium in order to provide technical support for its engineering applications. Oxidation kinetics of uranium and its alloys in oxidizing atmospheres are in good agreement with those in the literature. In addition to the traditional techniques, non-traditional methods have been applied for oxidation kinetics of uranium, and it has been verified that spectroscopic ellipsometry and X-ray diffraction are effective and nondestructive tools for in situ kinetic studies. The inhibition efficiency of oxidizing gas impurities on uranium hydrogenation is found to follow the order CO{sub 2} > CO > O{sub 2}, and the broadening of XPS shoulders with temperature in depth profile of hydrogenated uranium surface is discussed, which is not mentioned in the literature. Significant progress on surface chemistry of alloyed uranium (U-Nb and U-Ti) in hydrogen atmosphere is reported, and it is revealed that the hydrating nucleation and subsequent growth of alloyed uranium are closely connected with the surface states, underlying metal matrix, and it is microstructure-dependent. In this review, the recent advances in uranium surface chemistry in China, published so far mostly in Chinese language, are briefly summarized. Suggestions for further study are made. (orig.)

  17. Recent advances in study of uranium surface chemistry in China

    International Nuclear Information System (INIS)

    Luo, Lizhu; Lai, Xinchun; Wang, Xiaolin

    2014-01-01

    Uranium is very important in nuclear energy industry; however, uranium and its alloys corrode seriously in various atmospheres because of their chemical reactivities. In China, continuous investigations focused on surface chemistry have been carried out for a thorough understanding of uranium in order to provide technical support for its engineering applications. Oxidation kinetics of uranium and its alloys in oxidizing atmospheres are in good agreement with those in the literature. In addition to the traditional techniques, non-traditional methods have been applied for oxidation kinetics of uranium, and it has been verified that spectroscopic ellipsometry and X-ray diffraction are effective and nondestructive tools for in situ kinetic studies. The inhibition efficiency of oxidizing gas impurities on uranium hydrogenation is found to follow the order CO 2 > CO > O 2 , and the broadening of XPS shoulders with temperature in depth profile of hydrogenated uranium surface is discussed, which is not mentioned in the literature. Significant progress on surface chemistry of alloyed uranium (U-Nb and U-Ti) in hydrogen atmosphere is reported, and it is revealed that the hydrating nucleation and subsequent growth of alloyed uranium are closely connected with the surface states, underlying metal matrix, and it is microstructure-dependent. In this review, the recent advances in uranium surface chemistry in China, published so far mostly in Chinese language, are briefly summarized. Suggestions for further study are made. (orig.)

  18. Effect of electrical discharge machining on uranium-0.75 titanium and tungsten-3.5 nickel-1.5 iron alloys

    International Nuclear Information System (INIS)

    Anderson, R.C.

    1976-06-01

    It was found that U--0.75 Ti alloy cracked if the EDM parameters were out of control, and precipitation of carbides adjacent to the EDM surface took place during subsequent solution quenching. Cracks form in the ''recast'' layer when solution-quenched U--0.75 Ti alloy undergoes EDM, and the cracks propagated during subsequent nickel plating. If the recast layer was removed prior to nickel plating, only a slight loss in strength resulted, compared to conventional machining. W--3.5 Ni--1.5 Fe alloy also sustained some surface damage during EDM and also experienced a small loss in strength compared to conventionally machined material. 12 figures, 4 tables

  19. REGENERATION OF FISSION-PRODUCT-CONTAINING MAGNESIUM-THORIUM ALLOYS

    Science.gov (United States)

    Chiotti, P.

    1964-02-01

    A process of regenerating a magnesium-thorium alloy contaminated with fission products, protactinium, and uranium is presented. A molten mixture of KCl--LiCl-MgCl/sub 2/ is added to the molten alloy whereby the alkali, alkaline parth, and rare earth fission products (including yttrium) and some of the thorium and uranium are chlorinated and

  20. 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)

  1. Method of fabricating a uranium-bearing foil

    Science.gov (United States)

    Gooch, Jackie G [Seymour, TN; DeMint, Amy L [Kingston, TN

    2012-04-24

    Methods of fabricating a uranium-bearing foil are described. The foil may be substantially pure uranium, or may be a uranium alloy such as a uranium-molybdenum alloy. The method typically includes a series of hot rolling operations on a cast plate material to form a thin sheet. These hot rolling operations are typically performed using a process where each pass reduces the thickness of the plate by a substantially constant percentage. The sheet is typically then annealed and then cooled. The process typically concludes with a series of cold rolling passes where each pass reduces the thickness of the plate by a substantially constant thickness amount to form the foil.

  2. Direct reduction of uranium oxide(U3O8) by Li metal and U-metal(Fe, Ni) alloy formation in molten LiCl medium

    International Nuclear Information System (INIS)

    Cho, Young Hwan; Kim, Tack Jin; Choi, In Kyu; Kim, Won Ho; Jee, Kwang Yong

    2004-01-01

    Molten salt based electrochemical processes are proposed as a promising method for the future nuclear programs and more specifically for spent fuel processing. The lithium reduction has been introduced to convert actinide oxides into corresponding actinide metal by using lithium metal as a reductant in molten LiCl medium. We have applied similar lab-scale experiments to reduce uranium oxide in an effort to gain additional information on rates and mechanisms

  3. Use of vacuum in processing of uranium

    International Nuclear Information System (INIS)

    Saify, M.T.; Rai, C.B.; Singh, S.P.; Singh, R.P.

    2003-01-01

    Full text: Natural uranium in the form of metal and alloys with suitable heat treatment are being used as fuel in research and some of the power reactors. The fuel is required to satisfy the purity specification from the criteria of neutron economy, corrosion resistance and fabricability. Uranium and its alloys fall under the category of reactive materials. They readily react with atmospheric air to form oxides. If molten uranium is exposed to atmosphere, it reacts violently with atmospheric gases and moisture, leading to explosion in extreme cases. Hence, protective inert atmosphere or high vacuum is required in processing of the materials especially during the melting and casting operation. Vacuum is preferred for melting and remelting of metals and alloys to remove the gaseous and high volatile impurities, to improve the mechanical properties of the material. Also, under vacuum sound castings are produced for further processing by mechanical working or use in casting forms. The addition of reactive alloying elements in uranium is efficiently carried out under vacuum. The paper highlights vacuum systems deployed and applications of vacuum in various operations involved in the processing of uranium and its alloys

  4. Czechoslovak uranium

    International Nuclear Information System (INIS)

    Pluskal, O.

    1992-01-01

    Data and knowledge related to the prospecting, mining, processing and export of uranium ores in Czechoslovakia are presented. In the years between 1945 and January 1, 1991, 98,461.1 t of uranium were extracted. In the period 1965-1990 the uranium industry was subsidized from the state budget to a total of 38.5 billion CSK. The subsidies were put into extraction, investments and geologic prospecting; the latter was at first, ie. till 1960 financed by the former USSR, later on the two parties shared costs on a 1:1 basis. Since 1981 the prospecting has been entirely financed from the Czechoslovak state budget. On Czechoslovak territory uranium has been extracted from deposits which may be classified as vein-type deposits, deposits in uranium-bearing sandstones and deposits connected with weathering processes. The future of mining, however, is almost exclusively being connected with deposits in uranium-bearing sandstones. A brief description and characteristic is given of all uranium deposits on Czechoslovak territory, and the organization of uranium mining in Czechoslovakia is described as is the approach used in the world to evaluate uranium deposits; uranium prices and actual resources are also given. (Z.S.) 3 figs

  5. Controlled Environment Specimen Transfer

    DEFF Research Database (Denmark)

    Damsgaard, Christian Danvad; Zandbergen, Henny W.; Hansen, Thomas Willum

    2014-01-01

    an environmental transmission electron microscope to an in situ X-ray diffractometer through a dedicated transmission electron microscope specimen transfer holder, capable of sealing the specimen in a gaseous environment at elevated temperatures. Two catalyst material systems have been investigated; Cu/ZnO/Al2O3...... transferred in a reactive environment to the environmental transmission electron microscope where further analysis on the local scale were conducted. The Co/Al2O3 catalyst was reduced in the environmental microscope and successfully kept reduced outside the microscope in a reactive environment. The in situ......Specimen transfer under controlled environment conditions, such as temperature, pressure, and gas composition, is necessary to conduct successive complementary in situ characterization of materials sensitive to ambient conditions. The in situ transfer concept is introduced by linking...

  6. Reducing emissions from uranium dissolving

    International Nuclear Information System (INIS)

    Griffith, W.L.; Compere, A.L.; Huxtable, W.P.; Googin, J.M.

    1992-10-01

    This study was designed to assess the feasibility of decreasing NO x emissions from the current uranium alloy scrap tray dissolving facility. In the current process, uranium scrap is dissolved in boiling nitric acid in shallow stainless-steel trays. As scrap dissolves, more metal and more nitric acid are added to the tray by operating personnel. Safe geometry is assured by keeping liquid level at or below 5 cm, the depth of a safe infinite slab. The accountability batch control system provides additional protection against criticality. Both uranium and uranium alloys are dissolved. Nitric acid is recovered from the vapors for reuse. Metal nitrates are sent to uranium recovery. Brown NO x fumes evolved during dissolving have occasionally resulted in a visible plume from the trays. The fuming is most noticeable during startup and after addition of fresh acid to a tray. Present environmental regulations are expected to require control of brown NO x emissions. A detailed review of the literature, indicated the feasibility of slightly altering process chemistry to favor the production of NO 2 which can be scrubbed and recycled as nitric acid. Methods for controlling the process to manage offgas product distribution and to minimize chemical reaction hazards were also considered

  7. Study of transformations by annealing of the body. Centred cubic {gamma} phase of uranium-molybdenum alloys; Etude des transformations par revenu de la phase {gamma} cubique centree des alliages uranium-molybdene

    Energy Technology Data Exchange (ETDEWEB)

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

    1959-06-15

    By annealing at different temperatures, we have studied the transformations of the body centred cubic {gamma} phase for two alloys containing 6 and 10 per cent molybdenum by weight respectively. There is a return to the equilibrium state by formation of the stable {alpha} orthorhombic and {epsilon} ordered tetragonal phases, following two types of reaction: - pearlite transformation by nucleation and growth from the grain boundaries, preponderant when the annealing takes place at temperature above 400 deg. C, and identical for the two types of alloys. This reaction has already been studied by numerous authors, who have constructed the corresponding TTT curves, - transformation inside the grains of the quenched solid solution when annealing takes place at 400 deg. C or below: 6 per cent alloy - precipitation of fine a phase particles, followed by progressive ordering of the solid solution enriched in molybdenum, 10 per cent alloy - formation of small ordered regions and then a fine a phase precipitate. In the course of this work we have paid particular attention to the study of intragranular reactions after low-temperature annealing, the reactions involved in this case not having been explained up to the present. The {gamma} phase transformation has been studied by means of three techniques: micrography - microhardness tests - X-ray diffraction. (author) [French] Nous avons etudie les transformations par revenu a differentes temperatures, de la phase {gamma} cubique centree des alliages U-Mo trempes, pour deux alliages a 6 et a 10 pour cent de molybdene en poids. Il y a retour a l'etat d'equilibre par formation des phases stables {alpha} orthorhombique et quadratique ordonnee, suivant deux types de reactions: - transformation perlitique par germination et croissance a partir des joints de grains, preponderante lorsque le recuit a lieu a temperature superieure a 400 deg. C, et identique pour les deux types d'alliages. Cette reaction a deja ete etudiee par de nombreux

  8. Study of uranium - 20 Wt per cent plutonium-niobium alloys (1963); Etude d'alliages U-Pu-Nb a 20 pour cent en poids de plutonium (1963)

    Energy Technology Data Exchange (ETDEWEB)

    Abgrall, J; Barthelemy, P; Boucher, R [Commissariat a l' Energie Atomique, Fontenay-aux-Roses (France). Centre d' Etudes Nucleaires

    1963-07-01

    U-Pu-Nb alloys containing 20 wt per cent Pu and 10 - 20 - 30 - 40 - 50 or 60 wt per cent Nb have been studied principally to determine the feasibility of their use as fuel element. The fabrication, casting and homogenisation presented certain difficulties due specially to niobium. The transformation temperatures, thermal expansion coefficients and nature of phases have been determined by thermal analysis, dilatometry, micrography and X Rays diffraction. For similar compositions, U-Pu-Mo and U-Pu-Nb alloys have many common points concerning the presence of zeta phase (up to 40 wt per cent Nb), the coefficients of expansion, the good behaviour during thermal cycling and the good resistance to air oxidation in spite of zeta phase. In consequence, irradiation tests in EL{sub 3} reactor (Saclay) will be carried out in the near future. (authors) [French] Les alliages a 20 pour cent de plutonium, 10 - 20 - 30 - 40 - 50 - 60 pour cent de niobium et le complement en uranium ont ete etudies du point de vue de leur possibilite d'emploi comme combustible. Les problemes d'elaboration, de mise en forme et d'homogeneisation sont presentes. Ils sont relativement delicats. On a determine par analyse thermique, dilatometrie, micrographie et diffraction des rayons X les temperatures de transformation a l'etat solide, les coefficients de dilatation et la nature des phases. Pour des teneurs analogues, on retrouve de nombreux points communs avec les alliages U-Pu-Mo: presence de la phase zeta des U-Pu a temperature moyenne, coefficients de dilatation analogues, bonne tenue en cyclage thermique et bonne resistance a l'oxydation dans l'air malgre la presence de la phase zeta. Des essais d'irradiation dans EL{sub 3} vont etre entrepris. (auteurs)

  9. Study of uranium - 20 Wt per cent plutonium-niobium alloys (1963); Etude d'alliages U-Pu-Nb a 20 pour cent en poids de plutonium (1963)

    Energy Technology Data Exchange (ETDEWEB)

    Abgrall, J.; Barthelemy, P.; Boucher, R. [Commissariat a l' Energie Atomique, Fontenay-aux-Roses (France). Centre d' Etudes Nucleaires

    1963-07-01

    U-Pu-Nb alloys containing 20 wt per cent Pu and 10 - 20 - 30 - 40 - 50 or 60 wt per cent Nb have been studied principally to determine the feasibility of their use as fuel element. The fabrication, casting and homogenisation presented certain difficulties due specially to niobium. The transformation temperatures, thermal expansion coefficients and nature of phases have been determined by thermal analysis, dilatometry, micrography and X Rays diffraction. For similar compositions, U-Pu-Mo and U-Pu-Nb alloys have many common points concerning the presence of zeta phase (up to 40 wt per cent Nb), the coefficients of expansion, the good behaviour during thermal cycling and the good resistance to air oxidation in spite of zeta phase. In consequence, irradiation tests in EL{sub 3} reactor (Saclay) will be carried out in the near future. (authors) [French] Les alliages a 20 pour cent de plutonium, 10 - 20 - 30 - 40 - 50 - 60 pour cent de niobium et le complement en uranium ont ete etudies du point de vue de leur possibilite d'emploi comme combustible. Les problemes d'elaboration, de mise en forme et d'homogeneisation sont presentes. Ils sont relativement delicats. On a determine par analyse thermique, dilatometrie, micrographie et diffraction des rayons X les temperatures de transformation a l'etat solide, les coefficients de dilatation et la nature des phases. Pour des teneurs analogues, on retrouve de nombreux points communs avec les alliages U-Pu-Mo: presence de la phase zeta des U-Pu a temperature moyenne, coefficients de dilatation analogues, bonne tenue en cyclage thermique et bonne resistance a l'oxydation dans l'air malgre la presence de la phase zeta. Des essais d'irradiation dans EL{sub 3} vont etre entrepris. (auteurs)

  10. Report on the Synchrotron Characterization of U-Mo and U-Zr Alloys and the Modeling Results

    Energy Technology Data Exchange (ETDEWEB)

    Okuniewski, Maria A. [Purdue Univ., West Lafayette, IN (United States); Ganapathy, Varsha [Purdue Univ., West Lafayette, IN (United States); Hamilton, Brenden [Purdue Univ., West Lafayette, IN (United States); Cassutt, Paul [Purdue Univ., West Lafayette, IN (United States); Zhang, Fan [Purdue Univ., West Lafayette, IN (United States); Velaquez, Daniel [Illinois Inst. of Technology, Chicago, IL (United States); Seibert, Rachel [Illinois Inst. of Technology, Chicago, IL (United States); Terry, Jeff [Illinois Inst. of Technology, Chicago, IL (United States); Sprouster, David [Brookhaven National Lab. (BNL), Upton, NY (United States); Ecker, Lynne [Brookhaven National Lab. (BNL), Upton, NY (United States); Elbakhshwan, Mohamed [Brookhaven National Lab. (BNL), Upton, NY (United States)

    2016-09-01

    ABSTRACT Uranium-molybdenum (U-Mo) and uranium-zirconium (U-Zr) are two promising fuel candidates for nuclear transmutation reactors which burn long-lived minor actinides and fission products within fast spectrum reactors. The objectives of this research are centered on understanding the early stages of fuel performance through the examination of the irradiation induced microstructural changes in U-Zr and U-Mo alloys subjected to low neutron fluences. Specimens that were analyzed include those that were previously irradiated in the Advanced Test Reactor at INL. This most recent work has focused on a sub-set of the irradiated specimens, specifically U-Zr and U-Mo alloys that were irradiated to 0.01 dpa at temperatures ranging from (150-800oC). These specimens were analyzed with two types of synchrotron techniques, including X-ray absorption fine structure and X-ray diffraction. These techniques provide non-destructive microstructural analysis, including phase identification and quantitation, lattice parameters, crystallite sizes, as well as bonding, structure, and chemistry. Preliminary research has shown changes in the phase fractions, crystallite sizes, and lattice parameters as a function of irradiation and temperature. Future data analyses will continue to explore these microstructural changes.

  11. Uranium ores

    International Nuclear Information System (INIS)

    Poty, B.; Roux, J.

    1998-01-01

    The processing of uranium ores for uranium extraction and concentration is not much different than the processing of other metallic ores. However, thanks to its radioactive property, the prospecting of uranium ores can be performed using geophysical methods. Surface and sub-surface detection methods are a combination of radioactive measurement methods (radium, radon etc..) and classical mining and petroleum prospecting methods. Worldwide uranium prospecting has been more or less active during the last 50 years, but the rise of raw material and energy prices between 1970 and 1980 has incited several countries to develop their nuclear industry in order to diversify their resources and improve their energy independence. The result is a considerable increase of nuclear fuels demand between 1980 and 1990. This paper describes successively: the uranium prospecting methods (direct, indirect and methodology), the uranium deposits (economical definition, uranium ores, and deposits), the exploitation of uranium ores (use of radioactivity, radioprotection, effluents), the worldwide uranium resources (definition of the different categories and present day state of worldwide resources). (J.S.)

  12. Uranium market

    International Nuclear Information System (INIS)

    Rubini, L.A.; Asem, M.A.D.

    1990-01-01

    The historical development of the uranium market is present in two periods: The initial period 1947-1970 and from 1970 onwards, with the establishment of a commercial market. The world uranium requirements are derived from the corresponding forecast of nuclear generating capacity, with, particular emphasis to the brazilian requirements. The forecast of uranium production until the year 2000 is presented considering existing inventories and the already committed demand. The balance between production and requirements is analysed. Finally the types of contracts currently being used and the development of uranium prices in the world market are considered. (author)

  13. Uranium enrichment

    International Nuclear Information System (INIS)

    1990-01-01

    This report looks at the following issues: How much Soviet uranium ore and enriched uranium are imported into the United States and what is the extent to which utilities flag swap to disguise these purchases? What are the U.S.S.R.'s enriched uranium trading practices? To what extent are utilities required to return used fuel to the Soviet Union as part of the enriched uranium sales agreement? Why have U.S. utilities ended their contracts to buy enrichment services from DOE?

  14. Preserve specimens for reproducibility

    Czech Academy of Sciences Publication Activity Database

    Krell, F.-T.; Klimeš, Petr; Rocha, L. A.; Fikáček, M.; Miller, S. E.

    2016-01-01

    Roč. 539, č. 7628 (2016), s. 168 ISSN 0028-0836 Institutional support: RVO:60077344 Keywords : reproducibility * specimen * biodiversity Subject RIV: EH - Ecology, Behaviour Impact factor: 40.137, year: 2016 http://www.nature.com/nature/journal/v539/n7628/full/539168b.html

  15. Amorphous alloys in the U-Cr-V system

    International Nuclear Information System (INIS)

    Ray, R.; Musso, E.

    1979-01-01

    Amorphous uranium-chromium-vanadium alloys and a method of producing them are described. The uranium content of the alloys may vary between 60 and 80 atom percent, and chromium and vanadium between 0 and 40 atom percent, most particularly between 20 and 40 atom percent. A maximum of 10 atom percent of Cr or V may be replaced by other alloying elements, including metalloids and at least one transtion metal element. (LL)

  16. Steel alloys

    International Nuclear Information System (INIS)

    Bloom, E.E.; Stiegler, J.O.; Rowcliffe, A.F.; Leitnaker, J.M.

    1977-01-01

    The invention deals with a fuel element for fast breeder reactors. It consits essentially of a uranium oxide, nitride, or carbide or a mixture of these fuels with a plutonium or thorium oxide, nitride, or carbide. The fuel elements are coated with an austenitic stainless steel alloy. Inside the fuel elements, vacancies or small cavities are produced by neutron effects which causes the steel coating to swell. According to the invention, swelling is prevented by a modification of type 304, 316, 321, or 12 K 72HV commercial steels. They consist mainly of Fe, Cr, and Ni in a ratio determined by a temary diagram. They may also contain 1.8 to 2.3% by weight of Mo and a fraction of Si (0.7 to 2% by weight) and Ti(0.10 to 0.5% by weight) to prevent cavity formation. They are structurally modified by cold working. (IHOE) [de

  17. Corrosion of Al-7075 by uranium hexafluoride

    International Nuclear Information System (INIS)

    1989-01-01

    The results of the Al-7075 corrosion by uranium hexafluoride are presented in this work. The kinetic study shows that corrosion process occurs by two temperature dependent mechanism and that the alloy can be safely used up to 140 0 C. The corrosion film is formed by uranium oxifluoride with variable composition in depth. Two alternative corrosion models are proposed in order to explain the experimental results, as well as the tests taht will be carried out to confirm one of them [pt

  18. Uranium mining

    International Nuclear Information System (INIS)

    Lange, G.

    1975-01-01

    The winning of uranium ore is the first stage of the fuel cycle. The whole complex of questions to be considered when evaluating the profitability of an ore mine is shortly outlined, and the possible mining techniques are described. Some data on uranium mining in the western world are also given. (RB) [de

  19. Uranium enrichment

    International Nuclear Information System (INIS)

    1989-01-01

    GAO was asked to address several questions concerning a number of proposed uranium enrichment bills introduced during the 100th Congress. The bill would have restructured the Department of Energy's uranium enrichment program as a government corporation to allow it to compete more effectively in the domestic and international markets. Some of GAO's findings discussed are: uranium market experts believe and existing market models show that the proposed DOE purchase of a $750 million of uranium from domestic producers may not significantly increase production because of large producer-held inventories; excess uranium enrichment production capacity exists throughout the world; therefore, foreign producers are expected to compete heavily in the United States throughout the 1990s as utilities' contracts with DOE expire; and according to a 1988 agreement between DOE's Offices of Nuclear Energy and Defense Programs, enrichment decommissioning costs, estimated to total $3.6 billion for planning purposes, will be shared by the commercial enrichment program and the government

  20. Uranium resources

    International Nuclear Information System (INIS)

    1976-01-01

    This is a press release issued by the OECD on 9th March 1976. It is stated that the steep increases in demand for uranium foreseen in and beyond the 1980's, with doubling times of the order of six to seven years, will inevitably create formidable problems for the industry. Further substantial efforts will be needed in prospecting for new uranium reserves. Information is given in tabular or graphical form on the following: reasonably assured resources, country by country; uranium production capacities, country by country; world nuclear power growth; world annual uranium requirements; world annual separative requirements; world annual light water reactor fuel reprocessing requirements; distribution of reactor types (LWR, SGHWR, AGR, HWR, HJR, GG, FBR); and world fuel cycle capital requirements. The information is based on the latest report on Uranium Resources Production and Demand, jointly issued by the OECD's Nuclear Energy Agency (NEA) and the International Atomic Energy Agency. (U.K.)

  1. Standard guide for preparation of metallographic specimens

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2011-01-01

    1.1 The primary objective of metallographic examinations is to reveal the constituents and structure of metals and their alloys by means of a light optical or scanning electron microscope. In special cases, the objective of the examination may require the development of less detail than in other cases but, under nearly all conditions, the proper selection and preparation of the specimen is of major importance. Because of the diversity in available equipment and the wide variety of problems encountered, the following text presents for the guidance of the metallographer only those practices which experience has shown are generally satisfactory; it cannot and does not describe the variations in technique required to solve individual specimen preparation problems. Note 1—For a more extensive description of various metallographic techniques, refer to Samuels, L. E., Metallographic Polishing by Mechanical Methods, American Society for Metals (ASM) Metals Park, OH, 3rd Ed., 1982; Petzow, G., Metallographic Etchin...

  2. Influence of corrosive media on the mechanical resistance of the uranium-vanadium alloy containing 0.20% by weight. Hydrogen embrittlement

    International Nuclear Information System (INIS)

    Arnould-Laurent, Robert; Fidelle, J.-P.

    1976-10-01

    Tests were carried out on the alloy UV 0.2% in order to determine its limits of utilization. The alloy was shown to be sensitive to the following phenomena: intrinsic brittleness (FI) due to dissolved residual hydrogen from fabrication; cracking by stress corrosion (FCSC), possible in certain conditions owing to a passive but imperfect behavior of the metal surface (appearance of microcracks at the surface or corrosion pitting due to inadequate protection by the surface oxide layer); generalized stress accelerated corrosion (CGAC), of microscopic aspect similar to that observed for corrosion under H 2 gas. In practice these effects are obtained, singly or in combination, as follows: maintenance under dry argon - Fi; deformation tests to rupture in aqueous solutions (pH:2 to 14) or after exposure to a chlorinated solvent: FI + FCSC predominating. Below pH2 no stress corrosion; delayed fracture under damp air - at 80 deg and 100 deg C - FI + FCSC under high stresses, giving rise to short failure times (tr) - FI + CSC + CGAC with CGAC predominating under lower stresses, giving long failure times; at 20 and 60 deg C - FCSC + FI predominating. Under high stresses (leading to short failure times) the FCSC contribution increases with temperature [fr

  3. Uranium supply and demand

    Energy Technology Data Exchange (ETDEWEB)

    Spriggs, M J

    1976-01-01

    Papers were presented on the pattern of uranium production in South Africa; Australian uranium--will it ever become available; North American uranium resources, policies, prospects, and pricing; economic and political environment of the uranium mining industry; alternative sources of uranium supply; whither North American demand for uranium; and uranium demand and security of supply--a consumer's point of view. (LK)

  4. Behaviour of uranium under irradiation

    International Nuclear Information System (INIS)

    Adda, Y.; Mustelier, J.P.; Quere, Y.; Commissariat a l'Energie Atomique, Fontenay-aux-Roses

    1964-01-01

    The main results obtained in a study of the formation of defects caused in uranium by fission at low temperature are reported. By irradiation at 20 K. it was possible to determine the number of Frenkel pairs produced by one fission. An analysis of the curves giving the variations in electrical resistivity shows the size of the displacement spikes and the mechanism of defect creation due to fission. Irradiations at 77 K gave additional information, showing behaviour differences in the case of recrystallised and of cold worked uranium. The diffusion of rare gases was studied using metal-rare gas alloys obtained by electrical discharge, and samples of irradiated uranium. Simple diffusion is only responsible for the release of the rare gases under vacuum in cases where the rare gas content is very low (very slightly irradiated U). On the other hand when the concentration is higher (samples prepared by electrical discharge) the gas is given off by the formation, growth and coalescence of bubbles; the apparent diffusion coefficient is then quite different from the true coefficient and cannot be used in calculations on swelling. The various factors governing the phenomenon of simple diffusion were examined. It was shown in particular that a small addition of molybdenum could reduce the diffusion coefficient by a factor of 100. The precipitation of gas in uranium (Kr), in silver (Kr) and in Al-Li alloy (He) have been followed by measurement of the crystal parameter and of the electrical resistivity, and by electron microscope examination of thin films. The important part played by dislocations in the generation and growth of bubbles has been demonstrated, and it has been shown also that precipitation of bubbles on the dislocation lattice could block the development of recrystallisation. The results of these studies were compared with observations made on the swelling of uranium and uranium alloys U Mo and U Nb strongly irradiated between 400 and 700 C. In the case of Cubic

  5. Post-deformation examination of specimens subjected to SCC testing

    Energy Technology Data Exchange (ETDEWEB)

    Gussev, Maxim N. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Field, Kevin G. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Busby, Jeremy T. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Leonard, Keith J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2016-09-01

    This report details the results of post-radiation and post-deformation characterizations performed during FY 2015–FY 2016 on a subset of specimens that had previously been irradiated at high displacement per atom (dpa) damage doses. The specimens, made of commercial austenitic stainless steels and alloys, were subjected to stress-corrosion cracking tests (constant extension rate testing and crack growth testing) at the University of Michigan under conditions typical of nuclear power plants. After testing, the specimens were returned to Oak Ridge National Laboratory (ORNL) for further analysis and evaluation.

  6. Possibilities of using metal uranium fuel in heavy water reactors

    International Nuclear Information System (INIS)

    Djuric, B.; Mihajlovic, A.; Drobnjak, Dj.

    1965-11-01

    There are serious economic reasons for using metal uranium in heavy water reactors, because of its high density, i.e. high conversion factor, and low cost of fuel elements production. Most important disadvantages are swelling at high burnup and corrosion risk. Some design concepts and application of improved uranium obtained by alloying are promising for achievement of satisfactory stability of metal uranium under reactor operation conditions [sr

  7. Scrap uranium recycling via electron beam melting

    International Nuclear Information System (INIS)

    McKoon, R.

    1993-11-01

    A program is underway at the Lawrence Livermore National Laboratory (LLNL) to recycle scrap uranium metal. Currently, much of the material from forging and machining processes is considered radioactive waste and is disposed of by oxidation and encapsulation at significant cost. In the recycling process, uranium and uranium alloys in various forms will be processed by electron beam melting and continuously cast into ingots meeting applicable specifications for virgin material. Existing vacuum processing facilities at LLNL are in compliance with all current federal and state environmental, safety and health regulations for the electron beam melting and vaporization of uranium metal. One of these facilities has been retrofitted with an auxiliary electron beam gun system, water-cooled hearth, crucible and ingot puller to create an electron beam melt furnace. In this furnace, basic process R ampersand D on uranium recycling will be performed with the goal of eventual transfer of this technology to a production facility

  8. Uranium, depleted uranium, biological effects; Uranium, uranium appauvri, effets biologiques

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2001-07-01

    Physicists, chemists and biologists at the CEA are developing scientific programs on the properties and uses of ionizing radiation. Since the CEA was created in 1945, a great deal of research has been carried out on the properties of natural, enriched and depleted uranium in cooperation with university laboratories and CNRS. There is a great deal of available data about uranium; thousands of analyses have been published in international reviews over more than 40 years. This presentation on uranium is a very brief summary of all these studies. (author)

  9. Uranium toxicology

    International Nuclear Information System (INIS)

    Ferreyra, Mariana D.; Suarez Mendez, Sebastian

    1997-01-01

    In this paper are presented the methods and procedures optimized by the Nuclear Regulatory Authority (ARN) for the determination of: natural uranium mass, activity of enriched uranium in samples of: urine, mucus, filters, filter heads, rinsing waters and Pu in urine, adopted and in some cases adapted, by the Environmental Monitoring and Internal Dosimetry Laboratory. The analyzed material corresponded to biological and environmental samples belonging to the staff professionally exposed that work in plants of the nuclear fuel cycle. For a better comprehension of the activities of this laboratory, it is included a brief description of the uranium radiochemical toxicity and the limits internationally fixed to preserve the workers health

  10. PREPARATION OF ACTINIDE-ALUMINUM ALLOYS

    Science.gov (United States)

    Moore, R.H.

    1962-09-01

    BS>A process is given for preparing alloys of aluminum with plutonium, uranium, and/or thorium by chlorinating actinide oxide dissolved in molten alkali metal chloride with hydrochloric acid, chlorine, and/or phosgene, adding aluminum metal, and passing air and/or water vapor through the mass. Actinide metal is formed and alloyed with the aluminum. After cooling to solidification, the alloy is separated from the salt. (AEC)

  11. Influence of aggressive media on the mechanical behavior of the uranium--0.20 wt % vanadium alloy the role of hydrogen embrittlement

    International Nuclear Information System (INIS)

    Arnould-Laurent, R.

    The tests comprised tensile tests under constant load or up to the fracture point using cylindrical or flat, trapezoidal test pieces, tests in which disks were ruptured under gaseous pressure, and tenacity tests. The alloy was found to be sensitive to: (1) intrinsic brittleness (I.B.) due to dissolved residual hydrogen from the preparation stage. This manifested itself mainly by cracking at an elongation threshold of about 3 percent. (2) Cracking due to stress corrosion (S.C.C.) in the true sense, which is made possible under certain conditions by an imperfect passivation of the metal surface. The process is initiated either by the appearance of microcracks which appear at the surface, or by corrosion pits. (3) Generalized corrosion accelerated by the stress (S.A.C.), whose microscopic appearance is similar to that observed with corrosion under gaseous hydrogen. Below pH 2 there is no stress corrosion. Stress rupture tests in moist air at 80 and 100 0 C measure I.B. + S.C.C. under high stress, giving rise to short lifetimes. I.B. + S.C.C. + S.A.C., with S.A.C. predominant, occurs under lower stresses that give long lifetimes. Stress rupture tests measure at 20 and 60 0 C I.B. + S.C.C. with I.B. predominant. Under high stresses (short lifetimes) the magnitude of the S.C.C. component increases as the temperature increases. The most serious effects are those of S.A.C. at 80 and 100 0 C, and of I.B. at all temperatures. The way this alloy behaves can only be changed by an effective reduction in the quantity of residual hydrogen present, or by coatings that will in no case allow the ingress of hydrogen. 62 fig, 82 references, 15 tables

  12. NASA Biological Specimen Repository

    Science.gov (United States)

    McMonigal, K. A.; Pietrzyk, R. A.; Sams, C. F.; Johnson, M. A.

    2010-01-01

    The NASA Biological Specimen Repository (NBSR) was established in 2006 to collect, process, preserve and distribute spaceflight-related biological specimens from long duration ISS astronauts. This repository provides unique opportunities to study longitudinal changes in human physiology spanning may missions. The NBSR collects blood and urine samples from all participating ISS crewmembers who have provided informed consent. These biological samples are collected once before flight, during flight scheduled on flight days 15, 30, 60, 120 and within 2 weeks of landing. Postflight sessions are conducted 3 and 30 days after landing. The number of in-flight sessions is dependent on the duration of the mission. Specimens are maintained under optimal storage conditions in a manner that will maximize their integrity and viability for future research The repository operates under the authority of the NASA/JSC Committee for the Protection of Human Subjects to support scientific discovery that contributes to our fundamental knowledge in the area of human physiological changes and adaptation to a microgravity environment. The NBSR will institute guidelines for the solicitation, review and sample distribution process through establishment of the NBSR Advisory Board. The Advisory Board will be composed of representatives of all participating space agencies to evaluate each request from investigators for use of the samples. This process will be consistent with ethical principles, protection of crewmember confidentiality, prevailing laws and regulations, intellectual property policies, and consent form language. Operations supporting the NBSR are scheduled to continue until the end of U.S. presence on the ISS. Sample distribution is proposed to begin with selections on investigations beginning in 2017. The availability of the NBSR will contribute to the body of knowledge about the diverse factors of spaceflight on human physiology.

  13. Rotating specimen rack repair

    International Nuclear Information System (INIS)

    Miller, G.E.; Rogers, P.J.; Nabor, W.G.; Bair, H.

    1984-01-01

    In 1980, an operator at the UCI TRIGA Reactor noticed difficulties with the rotation of the specimen rack. Investigations showed that the drive bearing in the rack had failed and allowed the bearings to enter the rack. After some time of operation in static mode it was decided that installation of a bearing substitute - a graphite sleeve - would be undertaken. Procedures were written and approved for removal of the rack, fabrication and installation of the sleeve, and re-installation of the rack. This paper describes these procedures in some detail. Detailed drawings of the necessary parts may be obtained from the authors

  14. Method for thinning specimen

    Science.gov (United States)

    Follstaedt, David M.; Moran, Michael P.

    2005-03-15

    A method for thinning (such as in grinding and polishing) a material surface using an instrument means for moving an article with a discontinuous surface with an abrasive material dispersed between the material surface and the discontinuous surface where the discontinuous surface of the moving article provides an efficient means for maintaining contact of the abrasive with the material surface. When used to dimple specimens for microscopy analysis, a wheel with a surface that has been modified to produce a uniform or random discontinuous surface significantly improves the speed of the dimpling process without loss of quality of finish.

  15. Aluminum fin-stock alloys

    International Nuclear Information System (INIS)

    Gul, R.M.; Mutasher, F.

    2007-01-01

    Aluminum alloys have long been used in the production of heat exchanger fins. The comparative properties of the different alloys used for this purpose has not been an issue in the past, because of the significant thickness of the finstock material. However, in order to make fins lighter in weight, there is a growing demand for thinner finstock materials, which has emphasized the need for improved mechanical properties, thermal conductivity and corrosion resistance. The objective of this project is to determine the effect of iron, silicon and manganese percentage increment on the required mechanical properties for this application by analyzing four different aluminum alloys. The four selected aluminum alloys are 1100, 8011, 8079 and 8150, which are wrought non-heat treatable alloys with different amount of the above elements. Aluminum alloy 1100 serve as a control specimen, as it is commercially pure aluminum. The study also reports the effect of different annealing cycles on the mechanical properties of the selected alloys. Metallographic examination was also preformed to study the effect of annealing on the precipitate phases and the distribution of these phases for each alloy. The microstructure analysis of the aluminum alloys studied indicates that the precipitated phase in the case of aluminum alloys 1100 and 8079 is beta-FeAI3, while in 8011 it is a-alfa AIFeSi, and the aluminum alloy 8150 contains AI6(Mn,Fe) phase. The comparison of aluminum alloys 8011 and 8079 with aluminum alloy 1100 show that the addition of iron and silicon improves the percent elongation and reduces strength. The manganese addition increases the stability of mechanical properties along the annealing range as shown by the comparison of aluminum alloy 8150 with aluminum alloy 1100. Alloy 8150 show superior properties over the other alloys due to the reaction of iron and manganese, resulting in a preferable response to thermal treatment and improved mechanical properties. (author)

  16. Rossing uranium

    International Nuclear Information System (INIS)

    Anon.

    1979-01-01

    In this article the geology of the deposits of the Rossing uranium mine in Namibia is discussed. The planning of the open-pit mining, the blasting, drilling, handling and the equipment used for these processes are described

  17. Uranium, depleted uranium, biological effects

    International Nuclear Information System (INIS)

    2001-01-01

    Physicists, chemists and biologists at the CEA are developing scientific programs on the properties and uses of ionizing radiation. Since the CEA was created in 1945, a great deal of research has been carried out on the properties of natural, enriched and depleted uranium in cooperation with university laboratories and CNRS. There is a great deal of available data about uranium; thousands of analyses have been published in international reviews over more than 40 years. This presentation on uranium is a very brief summary of all these studies. (author)

  18. Uranium loans

    International Nuclear Information System (INIS)

    Anon.

    1990-01-01

    When NUEXCO was organized in 1968, its founders conceived of a business based on uranium loans. The concept was relatively straightforward; those who found themselves with excess supplies of uranium would deposit those excesses in NUEXCO's open-quotes bank,close quotes and those who found themselves temporarily short of uranium could borrow from the bank. The borrower would pay interest based on the quantity of uranium borrowed and the duration of the loan, and the bank would collect the interest, deduct its service fee for arranging the loan, and pay the balance to those whose deposits were borrowed. In fact, the original plan was to call the firm Nuclear Bank Corporation, until it was discovered that using the word open-quotes Bankclose quotes in the name would subject the firm to various US banking regulations. Thus, Nuclear Bank Corporation became Nuclear Exchange Corporation, which was later shortened to NUEXCO. Neither the nuclear fuel market nor NUEXCO's business developed quite as its founders had anticipated. From almost the very beginning, the brokerage of uranium purchases and sales became a more significant activity for NUEXCO than arranging uranium loans. Nevertheless, loan transactions have played an important role in the international nuclear fuel market, requiring the development of special knowledge and commercial techniques

  19. Determination of the Uranium Content of Aluminium Alloys; Determination de la Teneur en Uranium dans les Alliages a Base d'Aluminium; Opredelenie soderzhaniya urana v splavakh na osnove alyuminiya; Determinacion del Contenido de Uranio en las Aleaciones a Base de Aluminio

    Energy Technology Data Exchange (ETDEWEB)

    Gerard, J.; Van Hove, L. [S.A. Metallurgie et Mecanique Nucleaires Dessel (Belgium)

    1965-09-15

    Many materials testing reactors use as fuel an alloy of aluminium and enriched uranium. The amount of U{sup 235} in each fuel element must be known accurately. The techniques used to determine this are not only simple in principle but also varied; they include measurement of the alloy density, counting of the {gamma}-activity of the U{sup 235}, chemical analysis, determination of the isotopic content and evaluation of radiograph blackening. Unfortunately, all these methods possess more or less serious disadvantages when used for inspection on an industrial scale. The measurement of alloy density by Archimedes' method gives sufficiently accurate results if care is taken and the densities of the constituent metals are exactly known. The procedures employed, however, make this a very slow method. Moreover, the isotopic content must be known with great accuracy if the amount of U{sup 235} contained is to be determined. Counting the y-activity of the U{sup 235} yields a direct evaluation of this isotope, but the many parameters of a single-channel spectrometer require the use of accurately known standards and exceptionally stable counting facilities. Nevertheless, this method is in the authors' opinion the most suitable one for determinations on an industrial scale; and although the choice of standards requires care, and it is a slow and expensive matter to establish them, for any given level of production the choice is made once and for all, the only practical difficulty being to ensure the continued electronic stability of the installation. The other methods mentioned - chemical and isotopic analyses, and densitometry - are not. in common use. Measurement of the blackening of a radiograph is not sufficiently precise and does not permit the determination of uranium content with the required degree of accuracy. Densitometric examination of radiographs is nevertheless very useful, even indispensable, in assessing the homogeneity of an alloy or its mean content compared

  20. Orientational relationships between phases in the {gamma}{yields}{alpha} transformations for uranium-molybdenum alloys; Relations d'orientation entre phases dans les transformations {gamma}{yields}{alpha} des alliages uranium-molybdene

    Energy Technology Data Exchange (ETDEWEB)

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

    1966-04-01

    A crystallographic study has been made of the {gamma} {yields} {alpha} + {gamma} transformation in the alloy containing 3 per cent by weight of molybdenum using electronic micro-diffraction; it has been possible to establish the orientational relationships governing the germination of the {alpha} phase in the {gamma} phase. One finds: (111){gamma} // (100) {alpha}, (112-bar){gamma} // (010) {alpha}, (11-bar 0){gamma} // (001){alpha}. By choosing a monoclinic lattice containing the same number of atoms as the orthorhombic lattice for defining the {gamma} mother phase, the change in structure has been explained by adding a homogeneous (112-bar){gamma} [111]{gamma} shearing deformation to a heterogeneous deformation brought about by slipping of the atoms which are not situated at the nodes of this lattice. The identity of the orientation relationships {gamma}/{alpha} and {gamma}/{alpha}''b and the loss of coherence {gamma} /{alpha} as a function of temperature or of time lead to the conclusion that, in the range studied, the {gamma} {yields} {alpha} transformation begins with a martensitic process and continues by germination and growth. (author) [French] Une etude cristallographique de la transformation {gamma} {yields} {alpha} + {gamma} dans l'alliage {alpha} 3 pour cent en poids de Mo, effectuee par microdiffraction electronique a permis d'etablir les relations d'orientation regissant la germination de {alpha} dans {gamma}. On a: (111){gamma} // (100){alpha}, (112-bar){gamma} // (010){alpha}, (11-bar 0){gamma} // (001){alpha}. En choisissant pour decrire la phase mere {gamma} une maille monoclinique contenant le meme nombre d'atomes que la maille orthorhombique {alpha}, le changement de structure a ete explique en superposant a une deformation homogene par cisaillement (112-bar){gamma} [111]{gamma} une deformation heterogene par glissement des atomes non situes aux noeuds de cette maille. L identite des relations d'orientation {gamma}/{alpha} et {gamma} /{alpha

  1. SHEATHING URANIUM

    Science.gov (United States)

    Colbeck, E.W.

    1959-02-01

    A method is deseribed for forming a conveniently handled corrosion resistant U articlc comprising pouring molten U into an open-ended corrosion resistant metal eontainer such as Cu and its alloys, Al, or austenitic Ni stainless steel. The exposed surface of the cast U is covered with a metallic packing material such as a brazing flux consisting of Al-Si alloy. The container is sealed iii contact with substantially the entire exposed surface of the packing material. The article is then worked mechanically to reduce the cross section. l3651 A thorium--carbon alloy containing 0.1 to 0.5% by weight carbon, whieh is more resistant to water corrosion than pure thorium metal is presented. The alloy is prepared by fusing thorium metal with the desired amount of carbon at a temperature of about 1850 C. It is found that the carbon is present in the alloy as thorium monocarbide

  2. Conversion and Blending Facility highly enriched uranium to low enriched uranium as metal. Revision 1

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-07-05

    The mission of this Conversion and Blending Facility (CBF) will be to blend surplus HEU metal and alloy with depleted uranium metal to produce an LEU product. The primary emphasis of this blending operation will be to destroy the weapons capability of large, surplus stockpiles of HEU. The blended LEU product can only be made weapons capable again by the uranium enrichment process. The blended LEU will be produced as a waste suitable for storage or disposal.

  3. Conversion and Blending Facility highly enriched uranium to low enriched uranium as metal. Revision 1

    International Nuclear Information System (INIS)

    1995-01-01

    The mission of this Conversion and Blending Facility (CBF) will be to blend surplus HEU metal and alloy with depleted uranium metal to produce an LEU product. The primary emphasis of this blending operation will be to destroy the weapons capability of large, surplus stockpiles of HEU. The blended LEU product can only be made weapons capable again by the uranium enrichment process. The blended LEU will be produced as a waste suitable for storage or disposal

  4. Uranium decontamination of common metals by smelting, a review (handbook)

    International Nuclear Information System (INIS)

    Mautz, E.W.; Briggs, G.G.; Shaw, W.E.; Cavendish, J.H.

    1975-01-01

    The published and unpublished literature relating to the smelting of common metals scrap contaminated with uranium-bearing compounds has been searched and reviewed. In general, standard smelting practice produces ingots having a low uranium content, particularly for ferrous, nickel, and copper metals or alloys. Aluminum recovered from uranium contaminated scrap shows some decontamination by smelting but the uranium content is not as low as for other metals. Due to the heterogeneous nature and origin of scrap metals contaminated with uranium, information is frequently missing as to the extent of the initial contamination and the degree of decontamination obtained. The uranium content of the final cast ingots is generally all that is available. Results are summarized below by the primary composition of the uranium contaminated scrap metal. (U.S.)

  5. Proceedings of the JOWOG 22C (uranium) meeting

    Energy Technology Data Exchange (ETDEWEB)

    Nelson, T; Talaber, C; Wood, D H [eds.

    1987-01-01

    Lawrence Livermore National Laboratory was pleased to be host to the JOWOG 22C Meeting on June 9-11, 1987. This meeting was one of a continuing series on the subject of uranium and uranium alloys held between representatives of the United Kingdom and the United States under a treaty signed July 3, 1958. These, and similar meetings on other subjects, are controlled by the Department of Energy and the Joint Atomic Information Exchange Group (a combined agency of the Departments of Energy and Defense). The following topics were covered in the meeting: Use of Computers to Simulate Uranium; Corrosion and Chemical Stability; Superplasticity; Bonding, Corrosion, Etc.; Thermomechanical Properties and Fabrication; U-Ti Alloys; Uranium-Niobium Alloys; Physical Metallurgy and Testing; Miscellaneous Subjects; and Production and Facilities/Production Technology.

  6. Uranium extraction from gold-uranium ores

    Energy Technology Data Exchange (ETDEWEB)

    Laskorin, B.N.; Golynko, Z.Sh.

    1981-01-01

    The process of uranium extraction from gold-uranium ores in the South Africa is considered. Flowsheets of reprocessing gold-uranium conglomerates, pile processing and uranium extraction from the ores are presented. Continuous counter flow ion-exchange process of uranium extraction using strong-active or weak-active resins is noted to be the most perspective and economical one. The ion-exchange uranium separation with the succeeding extraction is also the perspective one.

  7. Studies on tempering at different temperatures of the beta phase retained by water quenching in uranium-chromium alloys containing from 0,37 to 4 atoms of chromium percent (1963); Etude du revenu a differentes temperatures de la phase beta retenue par trempe a l'eau dans les alliages uranium-chrome contenant de 0,37 a 4 atomes pour cent de chrome (1963)

    Energy Technology Data Exchange (ETDEWEB)

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

    1962-07-15

    The author made a systematic study of the annealing of the beta phase retained by water-quenching in uranium-chromium alloys of concentrations between 0.37 and 4 of chromium percent. It is shown that alloys containing less than 1 atom per cent are transformed at temperatures between room temperature and 250 deg. C according to a bainitic process involving activation energies of the order of 14,500 cal/mole. Alloys containing more than 1 at. per cent are transformed at temperature between 400 and 650 deg. C by way of a germination and growth process involving an activation energy of the order of 33,000 cal/mole. The limit of solubility of chromium in beta uranium plays a fundamental part in the transformations of the alloys. The TTT curves of beta {yields} alpha transformation were drawn by the use of a thermo-dilatometer of very low inertia. The transformation law may be expressed 1 x = exp. (kt){sup n}; x represents the degree of progression of the transformation, k a coefficient dependent on the temperature, and n an exponent depending only on the composition of the alloy. A micrographic and crystallographic study confirmed the results found by dilatometry; in particular it was possible to measure the progression rates of the transformation. (author) [French] L'auteur a fait une etude systematique du revenu de la phase beta retenue par trempe a l'eau dans les alliages uranium-chrome de teneurs comprises entre 0,37 et 4 atomes pour cent de chrome. Il a montre que les alliages qui contiennent moins de 1 atome pour cent de chrome se transforment aux temperatures comprises entre la temperature ordinaire et 250 deg. C selon un processus bainitique mettant en jeu des energies d'activatlon de l'ordre de 14500 cal/mole. Les alliages qui renferment plus de 1 atome pour cent de chrome se transforment aux temperatures comprises entre 400 et 650 deg. C suivant un processus de germination et croissance mettant en jeu une energie d'activation de l'ordre de -33000 cal/mole. La

  8. Uranium mining

    International Nuclear Information System (INIS)

    2008-01-01

    Full text: The economic and environmental sustainability of uranium mining has been analysed by Monash University researcher Dr Gavin Mudd in a paper that challenges the perception that uranium mining is an 'infinite quality source' that provides solutions to the world's demand for energy. Dr Mudd says information on the uranium industry touted by politicians and mining companies is not necessarily inaccurate, but it does not tell the whole story, being often just an average snapshot of the costs of uranium mining today without reflecting the escalating costs associated with the process in years to come. 'From a sustainability perspective, it is critical to evaluate accurately the true lifecycle costs of all forms of electricity production, especially with respect to greenhouse emissions, ' he says. 'For nuclear power, a significant proportion of greenhouse emissions are derived from the fuel supply, including uranium mining, milling, enrichment and fuel manufacture.' Dr Mudd found that financial and environmental costs escalate dramatically as the uranium ore is used. The deeper the mining process required to extract the ore, the higher the cost for mining companies, the greater the impact on the environment and the more resources needed to obtain the product. I t is clear that there is a strong sensitivity of energy and water consumption and greenhouse emissions to ore grade, and that ore grades are likely to continue to decline gradually in the medium to long term. These issues are critical to the current debate over nuclear power and greenhouse emissions, especially with respect to ascribing sustainability to such activities as uranium mining and milling. For example, mining at Roxby Downs is responsible for the emission of over one million tonnes of greenhouse gases per year and this could increase to four million tonnes if the mine is expanded.'

  9. Uranium enrichment

    International Nuclear Information System (INIS)

    Rae, H.K.; Melvin, J.G.

    1988-06-01

    Canada is the world's largest producer and exporter of uranium, most of which is enriched elsewhere for use as fuel in LWRs. The feasibility of a Canadian uranium-enrichment enterprise is therefore a perennial question. Recent developments in uranium-enrichment technology, and their likely impacts on separative work supply and demand, suggest an opportunity window for Canadian entry into this international market. The Canadian opportunity results from three particular impacts of the new technologies: 1) the bulk of the world's uranium-enrichment capacity is in gaseous diffusion plants which, because of their large requirements for electricity (more than 2000 kW·h per SWU), are vulnerable to competition from the new processes; 2) the decline in enrichment costs increases the economic incentive for the use of slightly-enriched uranium (SEU) fuel in CANDU reactors, thus creating a potential Canadian market; and 3) the new processes allow economic operation on a much smaller scale, which drastically reduces the investment required for market entry and is comparable with the potential Canadian SEU requirement. The opportunity is not open-ended. By the end of the century the enrichment supply industry will have adapted to the new processes and long-term customer/supplier relationships will have been established. In order to seize the opportunity, Canada must become a credible supplier during this century

  10. Vacuum fusion of uranium; Fusion de l'uranium sous vide

    Energy Technology Data Exchange (ETDEWEB)

    Stohr, J. A.

    1957-06-04

    After having outlined that vacuum fusion and moulding of uranium and of its alloys have some technical and economic benefits (vacuum operations avoid uranium oxidation and result in some purification; precision moulding avoids machining, chip production and chemical reprocessing of these chips; direct production of the desired shape is possible by precision moulding), this report presents the uranium fusion unit (its low pressure enclosure and pumping device, the crucible-mould assembly, and the MF supply device). The author describes the different steps of cast production, and briefly comments the obtained results.

  11. Uranium update

    International Nuclear Information System (INIS)

    Steane, R.

    1997-01-01

    This paper is about the current uranium mining situation, especially that in Saskatchewan. Canada has a unique advantage with the Saskatchewan uranium deposits. Making the most of this opportunity is important to Canada. The following is reviewed: project development and the time and capital it takes to bring a new project into production; the supply and demand situation to show where the future production fits into the world market; and our foreign competition and how we have to be careful not to lose our opportunity. (author)

  12. Exploration on relationship between uranium and organic materials in carbonate-siliceous pelite type uranium ore deposits

    International Nuclear Information System (INIS)

    Dong Yongjie

    1996-01-01

    The author determines the content of uranium and organic carbon of part specimen of surrounding rocks and ores, which sampled from carbonate and black shale type uranium deposits in Xiushui, Jiangxi Province, and Tongcheng, Hubei Province. According to the analytical operation regulations of organic materials, extraction and separation of chloroform pitch is carried out. Internal relationships between uranium and organic derivative is discussed. The conclusion shows that: (1) certain co-relationship between U and organic carbon and chloroform extract is detected; (2) evolutionary processes of organic materials in the exogenetic uranium deposits are not all the same; (3) non-hydrocarbon is closely related to uranium, so it can be regarded as indicator of uranium gathering in exogenetic uranium deposits

  13. Splitting tests on rock specimens

    Energy Technology Data Exchange (ETDEWEB)

    Davies, J D; Stagg, K G

    1970-01-01

    Splitting tests are described for a square-section sandstone specimens line loaded through steel or timber packings on the top face and supported on the bottom face either on similar packings (type A specimen) or directly on the lower platen plate of the testing machine (type B specimens). The stress distribution across the vertical central plane and the horizontal central plane were determined from a linear elastic finite element analysis for both types. Two solutions were obtained for the type B specimen: one assuming no friction between the base of the specimen and the platen plate and the other assuming no relative slip between the surfaces. Vertical and horizontal strains were measured at the center of the specimens for all loads up to failure.

  14. Janka hardness using nonstandard specimens

    Science.gov (United States)

    David W. Green; Marshall Begel; William Nelson

    2006-01-01

    Janka hardness determined on 1.5- by 3.5-in. specimens (2×4s) was found to be equivalent to that determined using the 2- by 2-in. specimen specified in ASTM D 143. Data are presented on the relationship between Janka hardness and the strength of clear wood. Analysis of historical data determined using standard specimens indicated no difference between side hardness...

  15. A melt refining method for uranium-contaminated aluminum

    International Nuclear Information System (INIS)

    Uda, T.; Iba, H.; Hanawa, K.

    1986-01-01

    Melt refining of uranium-contaminated aluminum which has been difficult to decontaminate because of the high reactivity of aluminum, was experimentally studied. Samples of contaminated aluminum and its alloys were melted after adding various halide fluxes at various melting temperatures and various melting times. Uranium concentration in the resulting ingots was determined. Effective flux compositions were mixtures of chlorides and fluorides, such as LiF, KCl, and BaCl 2 , at a fluoride/chloride mole ratio of 1 to 1.5. The removal of uranium from aluminum (the ''decontamination effect'') increased with decreasing melting temperature, but the time allowed for reaction had little influence. Pure aluminum was difficult to decontaminate from uranium; however, uranium could be removed from alloys containing magnesium. This was because the activity of the aluminum was decreased by formation of the intermetallic compound Al-Mg. With a flux of LiF-KCl-BaCl 2 and a temperature of 800 0 C, uranium added to give an initial concentration of 500 ppm was removed from a commercial alloy of aluminum, A5056, which contains 5% magnesium, to a final concentration of 0.6 ppm, which is near that in the initial aluminum alloy

  16. Uranium mining

    International Nuclear Information System (INIS)

    Cheeseman, E.W.

    1980-01-01

    The international uranium market appears to be currently over-supplied with a resultant softening in prices. Buyers on the international market are unhappy about some of the restrictions placed on sales by the government, and Canadian sales may suffer as a result. About 64 percent of Canada's shipments come from five operating Ontario mines, with the balance from Saskatchewan. Several other properties will be producing within the next few years. In spite of the adverse effects of the Three Mile Island incident and the default by the T.V.A. of their contract, some 3 600 tonnes of new uranium sales were completed during the year. The price for uranium had stabilized at US $42 - $44 by mid 1979, but by early 1980 had softened somewhat. The year 1979 saw the completion of major environmental hearings in Ontario and Newfoundland and the start of the B.C. inquiry. Two more hearings are scheduled for Saskatchewan in 1980. The Elliot Lake uranium mining expansion hearings are reviewed, as are other recent hearings. In the production of uranium for nuclear fuel cycle, environmental matters are of major concern to the industry, the public and to governments. Research is being conducted to determine the most effective method for removing radium from tailings area effluents. Very stringent criteria are being drawn up by the regulatory agencies that must be met by the industry in order to obtain an operating licence from the AECB. These criteria cover seepages from the tailings basin and through the tailings retention dam, seismic stability, and both short and long term management of the tailings waste management area. (auth)

  17. Uranium industry annual 1996

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-04-01

    The Uranium Industry Annual 1996 (UIA 1996) provides current statistical data on the US uranium industry`s activities relating to uranium raw materials and uranium marketing. The UIA 1996 is prepared for use by the Congress, Federal and State agencies, the uranium and nuclear electric utility industries, and the public. Data on uranium raw materials activities for 1987 through 1996 including exploration activities and expenditures, EIA-estimated reserves, mine production of uranium, production of uranium concentrate, and industry employment are presented in Chapter 1. Data on uranium marketing activities for 1994 through 2006, including purchases of uranium and enrichment services, enrichment feed deliveries, uranium fuel assemblies, filled and unfilled market requirements, uranium imports and exports, and uranium inventories are shown in Chapter 2. A feature article, The Role of Thorium in Nuclear Energy, is included. 24 figs., 56 tabs.

  18. Uranium industry annual 1996

    International Nuclear Information System (INIS)

    1997-04-01

    The Uranium Industry Annual 1996 (UIA 1996) provides current statistical data on the US uranium industry's activities relating to uranium raw materials and uranium marketing. The UIA 1996 is prepared for use by the Congress, Federal and State agencies, the uranium and nuclear electric utility industries, and the public. Data on uranium raw materials activities for 1987 through 1996 including exploration activities and expenditures, EIA-estimated reserves, mine production of uranium, production of uranium concentrate, and industry employment are presented in Chapter 1. Data on uranium marketing activities for 1994 through 2006, including purchases of uranium and enrichment services, enrichment feed deliveries, uranium fuel assemblies, filled and unfilled market requirements, uranium imports and exports, and uranium inventories are shown in Chapter 2. A feature article, The Role of Thorium in Nuclear Energy, is included. 24 figs., 56 tabs

  19. Uranium industry annual, 1991

    International Nuclear Information System (INIS)

    1992-10-01

    In the Uranium Industry Annual 1991, data on uranium raw materials activities including exploration activities and expenditures, resources and reserves, mine production of uranium, production of uranium concentrate, and industry employment are presented in Chapter 1. Data on uranium marketing activities including domestic uranium purchases, commitments by utilities, procurement arrangements, uranium imports under purchase contracts and exports, deliveries to enrichment suppliers, inventories, secondary market activities, utility market requirements, and uranium for sale by domestic suppliers are presented in Chapter 2. A feature article entitled ''The Uranium Industry of the Commonwealth of Independent States'' is included in this report

  20. Ultrasonic Inspection following Heat Treatment of Uranium Alloys; Controle des Traitements Thermiques d'Alliage d'Uranium par Ultrasons; Kontrol' termicheskoj obrabotki uranovykh splavov s pomoshch'yu ul'trazvuka; Control Ultrasonico de los Tratamientos Termicos de Aleaciones de Uranio

    Energy Technology Data Exchange (ETDEWEB)

    Destribats, Marie-Therese; Cherpentier, C.; Papezik, F.; Pigeon, M. [Centre d' Etudes Nucleaires Desaclay (France)

    1965-10-15

    To improve the behaviour of low uranium alloys in reactors it is often necessary to reduce grain size by heat treatment. It has proved essential to provide for inspection of the whole element and the entire output in order to discover the exact quality of the fuel used. This inspection cannot be made by micrography because of the time required and the fact that the data obtained are incomplete. The inspection system adopted is based on the principle of absorption of ultrasonic waves by materials. This absorption depends on the structure of the medium. If {lambda} is small in relation to grain size G, absorption is low; whereas if G is of the order of {lambda}/2, absorption is very high. The tests were made first in air, using the multiple-echo system, then by measuring the height of the first echo, and finally by transmission in water, the height of the transmitted echo being compared with that of the initial signal. In industrial use, the amplitude of the echo transmitted by the material is compared with the echo obtained from a standard of the same characteristics and shape. Inspection takes place in a special machine in which the materials are rotated by rollers and adjustable transducers move over the element. The helicoidal scanning is carried out with a pitch of less than 5 mm. The ultrasonic generator includes a control system ensuring a constant reference echo. The paper quotes a series of records showing the results obtained with various alloys and in particular the faults observed in elements treated by induction upon linear displacement. The arrangement can detect faulty treatment zones of less than 1 cm{sup 2}. The system is at present used to inspect all low alloy uranium fuels of the G2, EL3, EDF1, EDF2 and INCA reactors, i.e. rods and tubes with diameters between 20 and 95 mm. (author) [French] Afin d'obtenir une meilleure tenue des alliages d'uranium faiblement allies dans les reacteurs, un affinage du grain par traitements thermiques est souvent

  1. Instrumented impact testing machine with reduced specimen oscillation effects

    International Nuclear Information System (INIS)

    Rintamaa, R.; Rahka, K.; Wallin, K.

    1984-07-01

    Owing to small and inexpensive specimens the Charpy impact test is widely used in quality control and alloy development. Limitations in power reactor survellance capsules it is also widely used for safety analysis purposes. Instrumenting the tup and computerizing data acquisition, makes dynamic fracture mechanics data measurement possible and convenient. However, the dynamic effects (inertia forces, specimen oscillations) in the impact test cause inaccuracies in the recorded load-time diagram and hence diminish the reliability of the calculated dynamic fracture mechanics parameters. To decrease inaccuracies a new pendulum type of instrumented impact test apparatus has been developed and constructed in the Metals Laboratory of the Technical Research Centre of Finland. This tester is based on a new principle involving inverted test geometry. The purpose of the geometry inversion is to reduce inertia load and specimen oscillation effects. Further, the new impact tester has some other novel features: e.g. the available initia impact energy is about double compared to the conventional standard (300 J) impact tester allowing the use of larger (10 x 20 x 110 mm) bend specimens than normal Charpy specimens. Also, the rotation asix in the three point bending is nearly stationary making COD-measurements possible. An experimental test series is described in which the inertia effects and specimen oscillations are compared in the conventional and new impact tester utilizing Charpy V-notch specimens. Comparison of the two test geometries is also made with the aid of an analytical model using finite element method (FEM) analysis. (author)

  2. FORMING TUBES AND RODS OF URANIUM METAL BY EXTRUSION

    Science.gov (United States)

    Creutz, E.C.

    1959-01-27

    A method and apparatus are presented for the extrusion of uranium metal. Since uranium is very brittle if worked in the beta phase, it is desirable to extrude it in the gamma phase. However, in the gamma temperature range thc uranium will alloy with the metal of the extrusion dic, and is readily oxidized to a great degree. According to this patent, uranium extrusion in thc ganmma phase may be safely carried out by preheating a billet of uranium in an inert atmosphere to a trmperature between 780 C and 1100 C. The heated billet is then placed in an extrusion apparatus having dies which have been maintained at an elevated temperature for a sufficient length of time to produce an oxide film, and placing a copper disc between the uranium billet and the die.

  3. Miniature specimen technology for postirradiation fatigue crack growth testing

    International Nuclear Information System (INIS)

    Mervyn, D.A.; Ermi, A.M.

    1979-01-01

    Current magnetic fusion reactor design concepts require that the fatigue behavior of candidate first wall materials be characterized. Fatigue crack growth may, in fact, be the design limiting factor in these cyclic reactor concepts given the inevitable presence of crack-like flaws in fabricated sheet structures. Miniature specimen technology has been developed to provide the large data base necessary to characterize irradiation effects on the fatigue crack growth behavior. An electrical potential method of measuring crack growth rates is employed on miniature center-cracked-tension specimens (1.27 cm x 2.54 cm x 0.061 cm). Results of a baseline study on 20% cold-worked 316 stainless steel, which was tested in an in-cell prototypic fatigue machine, are presented. The miniature fatigue machine is designed for low cost, on-line, real time testing of irradiated fusion candidate alloys. It will enable large scale characterization and development of candidate first wall alloys

  4. Processing and properties of Nb-Ti-based alloys

    International Nuclear Information System (INIS)

    Sikka, V.K.; Viswanathan, S.

    1992-01-01

    The processing characteristics, tensile properties, and oxidation response of two Nb-Ti-Al-Cr alloys were investigated. One creep test at 650 C and 172 MPa was conducted on the base alloy which contained 40Nb-40Ti-10Al-10Cr. A second alloy was modified with 0.11 at. % carbon and 0.07 at. % yttrium. Alloys were arc melted in a chamber backfilled with argon, drop cast into a water-cooled copper mold, and cold rolled to obtain a 0.8-mm sheet. The sheet was annealed at 1,100 C for 0.5 h. Longitudinal tensile specimens and oxidation specimens were obtained for both the base alloy and the modified alloy. Tensile properties were obtained for the base alloy at room temperature, 400, 600, 700, 800, 900, and 1,000 C, and for the modified alloy at room temperature, 400, 600, 700, and 800 C. Oxidation tests on the base alloy and modified alloy, as measured by weight change, were carried out at 600, 700, 800, and 900 C. Both the base alloy and the modified alloy were extremely ductile and were cold rolled to the final sheet thickness of 0.8 mm without an intermediate anneal. The modified alloy exhibited some edge cracking during cold during cold rolling. Both alloys recrystallized at the end of a 0.5-h annealing treatment. The alloys exhibited moderate strength and oxidation resistance below 600 C, similar to the results of alloys reported in the literature

  5. Irradiation of copper alloys in FFTF

    International Nuclear Information System (INIS)

    Brager, H.R.; Garner, F.A.

    1984-01-01

    Nine copper-base alloys in thirteen material conditions have been inserted into the MOTA-18 experiment for irradiation in FFTF at approx.450 0 C. The alloy Ni-1.9Be is also included in this experiment, which includes both TEM disks and miniature tensile specimens

  6. Fundamental irradiation studies on vanadium alloys

    International Nuclear Information System (INIS)

    Loomis, B.A.; Garner, F.A.; Ermi, A.M.

    1985-05-01

    A joint experiment on the irradiation response of simple vanadium alloys has been initiated under the auspices of the DAFS and BES progams. Specimen fabrication is nearly complete and the alloys are expected to be irradiated in lithium in FFTF-MOTA Cycles 7 and 8

  7. Depleted-Uranium Weapons: the Whys and Wherefores

    OpenAIRE

    Gsponer, Andre

    2003-01-01

    The only military application in which depleted-uranium (DU) alloys out-perform present-day tungsten alloys is long-rod penetration into a main battle-tank's armor. However, this advantage is only on the order of 10%, and it disappears when the comparison is made in terms of actual lethality of complete anti-tank systems instead of laboratory-type steel penetration capability. Therefore, new micro- and nano-engineered tungsten alloys may soon out-perform DU alloys, enabling the production of ...

  8. Modification of the hydriding of uranium using ion implantation

    International Nuclear Information System (INIS)

    Musket, R.G.; Robinson-Weis, G.; Patterson, R.G.

    1983-01-01

    The hydriding of depleted uranium at 76 Torr hydrogen and 130 0 C has been significantly reduced by implantation of oxygen ions. The high-dose implanted specimens had incubation times for the initiation of the reaction after exposure to hydrogen that exceeded those of the nonimplanted specimens by more than a factor of eight. Furthermore, the nonimplanted specimens consumed enough hydrogen to cause macroscopic flaking of essentially the entire surface in times much less than the incubation time for the high-dose implanted specimens. In contrast, the ion-implanted specimens reacted only at isolated spots with the major fraction of the surface area unaffected by the hydrogen exposure

  9. Uranium - what role

    International Nuclear Information System (INIS)

    Grey, T.; Gaul, J.; Crooks, P.; Robotham, R.

    1980-01-01

    Opposing viewpoints on the future role of uranium are presented. Topics covered include the Australian Government's uranium policy, the status of nuclear power around the world, Australia's role as a uranium exporter and problems facing the nuclear industry

  10. Brazilian uranium exploration program

    International Nuclear Information System (INIS)

    Marques, J.P.M.

    1981-01-01

    General information on Brazilian Uranium Exploration Program, are presented. The mineralization processes of uranium depoits are described and the economic power of Brazil uranium reserves is evaluated. (M.C.K.) [pt

  11. Uranium enrichment

    International Nuclear Information System (INIS)

    1991-11-01

    This paper analyzes under four different scenarios the adequacy of a $500 million annual deposit into a fund to pay for the cost of cleaning up the Department of Energy's (DOE) three aging uranium enrichment plants. These plants are located in Oak Ridge, Tennessee; Paducah, Kentucky; and Portsmouth, Ohio. In summary the following was found: A fixed annual $500 million deposit made into a cleanup fund would not be adequate to cover total expected cleanup costs, nor would it be adequate to cover expected decontamination and decommissioning (D and D) costs. A $500 million annual deposit indexed to an inflation rate would likely be adequate to pay for all expected cleanup costs, including D and D costs, remedial action, and depleted uranium costs

  12. Uranium production

    International Nuclear Information System (INIS)

    Spriggs, M.

    1980-01-01

    The balance between uranium supply and demand is examined. Should new resources become necessary, some unconventional sources which could be considered include low-grade extensions to conventional deposits, certain types of intrusive rock, tuffs, and lake and sea-bed sediments. In addition there are large but very low grade deposits in carbonaceous shales, granites, and seawater. The possibility of recovery is discussed. Programmes of research into the feasibility of extraction of uranium from seawater, as a by-product from phosphoric acid production, and from copper leach solutions, are briefly discussed. Other possible sources are coal, old mine dumps and tailings, the latter being successfully exploited commercially in South Africa. The greatest constraints on increased development of U from lower grade sources are economics and environmental impact. It is concluded that apart from U as a by-product from phosphate, other sources are unlikely to contribute much to world requirements in the foreseeable future. (U.K.)

  13. Uranium enrichment

    International Nuclear Information System (INIS)

    1991-08-01

    This paper reports that in 1990 the Department of Energy began a two-year project to illustrate the technical and economic feasibility of a new uranium enrichment technology-the atomic vapor laser isotope separation (AVLIS) process. GAO believes that completing the AVLIS demonstration project will provide valuable information about the technical viability and cost of building an AVLIS plant and will keep future plant construction options open. However, Congress should be aware that DOE still needs to adequately demonstrate AVLIS with full-scale equipment and develop convincing cost projects. Program activities, such as the plant-licensing process, that must be completed before a plant is built, could take many years. Further, an updated and expanded uranium enrichment analysis will be needed before any decision is made about building an AVLIS plant. GAO, which has long supported legislation that would restructure DOE's uranium enrichment program as a government corporation, encourages DOE's goal of transferring AVLIS to the corporation. This could reduce the government's financial risk and help ensure that the decision to build an AVLIS plant is based on commercial concerns. DOE, however, has no alternative plans should the government corporation not be formed. Further, by curtailing a planned public access program, which would have given private firms an opportunity to learn about the technology during the demonstration project, DOE may limit its ability to transfer AVLIS to the private sector

  14. Derived enriched uranium market

    International Nuclear Information System (INIS)

    Rutkowski, E.

    1996-01-01

    The potential impact on the uranium market of highly enriched uranium from nuclear weapons dismantling in the Russian Federation and the USA is analyzed. Uranium supply, conversion, and enrichment factors are outlined for each country; inventories are also listed. The enrichment component and conversion components are expected to cause little disruption to uranium markets. The uranium component of Russian derived enriched uranium hexafluoride is unresolved; US legislation places constraints on its introduction into the US market

  15. Alloy materials

    Energy Technology Data Exchange (ETDEWEB)

    Hans Thieme, Cornelis Leo (Westborough, MA); Thompson, Elliott D. (Coventry, RI); Fritzemeier, Leslie G. (Acton, MA); Cameron, Robert D. (Franklin, MA); Siegal, Edward J. (Malden, MA)

    2002-01-01

    An alloy that contains at least two metals and can be used as a substrate for a superconductor is disclosed. The alloy can contain an oxide former. The alloy can have a biaxial or cube texture. The substrate can be used in a multilayer superconductor, which can further include one or more buffer layers disposed between the substrate and the superconductor material. The alloys can be made a by process that involves first rolling the alloy then annealing the alloy. A relatively large volume percentage of the alloy can be formed of grains having a biaxial or cube texture.

  16. Uranium industry annual, 1986

    International Nuclear Information System (INIS)

    1987-01-01

    Uranium industry data collected in the EIA-858 survey provide a comprehensive statistical characterization of annual activities of the industry and include some information about industry plans over the next several years. This report consists of two major sections. The first addresses uranium raw materials activities and covers the following topics: exploration activities and expenditures, resources and reserves, mine production of uranium, production of uranium concentrate, and industry employment. The second major section is concerned with the following uranium marketing activities: uranium purchase commitments, uranium prices, procurement arrangements, uranium imports and exports, enrichment services, inventories, secondary market activities utility market requirements and related topics

  17. Uranium Industry. Annual 1984

    International Nuclear Information System (INIS)

    Lawrence, M.S.S.

    1985-01-01

    This report provides a statistical description of activities of the US uranium industry during 1984 and includes a statistical profile of the status of the industry at the end of 1984. It is based on the results of an Energy Information Administration (EIA) survey entitled ''Uranium Industry Annual Survey'' (Form EIA-858). The principal findings of the survey are summarized under two headings - Uranium Raw Materials Activities and Uranium Marketing Activities. The first heading covers exploration and development, uranium resources, mine and mill production, and employment. The second heading covers uranium deliveries and delivery commitments, uranium prices, foreign trade in uranium, inventories, and other marketing activities. 32 figs., 48 tabs

  18. Evaluation of A-1 reactor heavy-water calandria specimens

    International Nuclear Information System (INIS)

    Brumovsky, M.

    1976-01-01

    Container chains with surveillance specimens were placed in two special channels of the core peripheral part to test changes in mechanical properties due to reactor operation of caisson tube material. The specimens were made from the caisson tube material and placed by eight pieces on the outer surface of the containers. The first removed specimens were tested for corrosion losses, tensile strength, and fractured surfaces were then assessed. The changes in strength properties were found to be similar in both base material and welded joints. The corrosion film on surveillance specimens did not practically affect strength properties nor ductility. It was found that the Al-Mg-Si alloy used for the heavy water vessel caisson tubes following stabilization annealing was fully stable at operating temperatures of up to 100 degC. Slio.ht changes in properties can be attributed to the effect of a high neutron dose. Thus, the high radiation and temperature stability of the alloy was confirmed. (O.K.)

  19. Shot peening influence on corrosion resistance of AE21 magnesium alloy.

    Science.gov (United States)

    2010-12-15

    "Evaluation of the electrochemical characteristics of the AE21 magnesium alloy is presented in the article. : The surfaces of tested alloys were treated by grinding and grinding followed by sodium bicarbonate shotpeening. : The specimens were evaluat...

  20. Characteristics of Film Formed on Alloy 600 and Alloy 690 in Water Containing lead

    International Nuclear Information System (INIS)

    Hwang Seong Sik; Lee, Deok Hyun; Kim, Hong Pyo; Kim, Joung Soo; Kim, Ju Yup

    1999-01-01

    Anodic polarization behaviors of Alloy 600 and Alloy 690 have been studied as a function of lead content in the solution of pH 4 and 10 at 90 .deg. C. As the amount of lead in the solution increased, critical current densities and passive current densities of Alloy 600 and Alloy 690 increased, while the breakdown potential of the alloys decreased. The high critical current density in the high lead solution was thought to come from the combination of an enhanced dissolution of constituents on the surface of the alloys by the lead and an anodic dissolution of metallic lead deposited on the surface of the specimens. The morphology of lead precipitated on the specimen after the anodic scan changed with the pH of solution: small irregular particles were precipitated on the surface of the specimen in the solution of pH 4, while the high density of regular sized particles was formed on it in the solution of pH 10.Pb was observed to enhance Cr depletion from the outer surface of Alloy 600 and Alloy 690 and also to increase the ratio of O 2- /OH - in the surface film formed in the high lead solution. The SCC resistance of Alloy 600 and Alloy 690 may have decreased due to the poor quality of the passive film formed and the enhanced oxygen evolution in the solution containing lead

  1. Uranium price reporting systems

    International Nuclear Information System (INIS)

    1987-09-01

    This report describes the systems for uranium price reporting currently available to the uranium industry. The report restricts itself to prices for U 3 O 8 natural uranium concentrates. Most purchases of natural uranium by utilities, and sales by producers, are conducted in this form. The bulk of uranium in electricity generation is enriched before use, and is converted to uranium hexafluoride, UF 6 , prior to enrichment. Some uranium is traded as UF 6 or as enriched uranium, particularly in the 'secondary' market. Prices for UF 6 and enriched uranium are not considered directly in this report. However, where transactions in UF 6 influence the reported price of U 3 O 8 this influence is taken into account. Unless otherwise indicated, the terms uranium and natural uranium used here refer exclusively to U 3 O 8 . (author)

  2. Uranium Industry Annual, 1992

    International Nuclear Information System (INIS)

    1993-01-01

    The Uranium Industry Annual provides current statistical data on the US uranium industry for the Congress, Federal and State agencies, the uranium and electric utility industries, and the public. The feature article, ''Decommissioning of US Conventional Uranium Production Centers,'' is included. Data on uranium raw materials activities including exploration activities and expenditures, resources and reserves, mine production of uranium, production of uranium concentrate, and industry employment are presented in Chapter 1. Data on uranium marketing activities including domestic uranium purchases, commitments by utilities, procurement arrangements, uranium imports under purchase contracts and exports, deliveries to enrichment suppliers, inventories, secondary market activities, utility market requirements, and uranium for sale by domestic suppliers are presented in Chapter 2

  3. Uranium Industry Annual, 1992

    Energy Technology Data Exchange (ETDEWEB)

    1993-10-28

    The Uranium Industry Annual provides current statistical data on the US uranium industry for the Congress, Federal and State agencies, the uranium and electric utility industries, and the public. The feature article, ``Decommissioning of US Conventional Uranium Production Centers,`` is included. Data on uranium raw materials activities including exploration activities and expenditures, resources and reserves, mine production of uranium, production of uranium concentrate, and industry employment are presented in Chapter 1. Data on uranium marketing activities including domestic uranium purchases, commitments by utilities, procurement arrangements, uranium imports under purchase contracts and exports, deliveries to enrichment suppliers, inventories, secondary market activities, utility market requirements, and uranium for sale by domestic suppliers are presented in Chapter 2.

  4. Automated controlled-potential coulometric determination of uranium

    International Nuclear Information System (INIS)

    Knight, C.H.; Clegg, D.E.; Wright, K.D.; Cassidy, R.M.

    1982-06-01

    A controlled-potential coulometer has been automated in our laboratory for routine determination of uranium in solution. The CRNL-designed automated system controls degassing, prereduction, and reduction of the sample. The final result is displayed on a digital coulometer readout. Manual and automated modes of operation are compared to show the precision and accuracy of the automated system. Results are also shown for the coulometric titration of typical uranium-aluminum alloy samples

  5. Screen-film specimen radiography

    International Nuclear Information System (INIS)

    Shepard, S.J.; Hogan, J.; Schreck, B.

    1990-01-01

    This paper reports on the reproducibility and quality of biopsy specimen radiographs, a unique phototimed cabinet x-ray system is being developed. The system utilizes specially modified Kodal Min-R cassettes and will be compatible with current mammographic films. Tube voltages are in the 14-20-kVp range with 0.1-1.0-second exposure times. A top-hat type compression device is used (1) to compress the specimen to uniform thickness, (2) to measure the specimen thickness and determine optimum kVp, and (3) to superimpose a grid over the specimen for identification of objects of radiographic interest. The phototiming circuit developed specifically for this purpose will be described along with the modified Min-R cassette. Characteristics of the generator and cabinet will also be described. Tests will be performed on phantoms to evaluate the system limitations

  6. f-band narrowing in uranium intermetallics

    International Nuclear Information System (INIS)

    Dunlap, B.D.; Litterst, F.J.; Malik, S.K.; Kierstead, H.A.; Crabtree, G.W.; Kwok, W.; Lam, D.J.; Mitchell, A.W.

    1987-01-01

    Although the discovery of heavy fermion behavior in uranium compounds has attracted a great deal of attention, relatively little work has been done which is sufficiently systematic to allow an assessment of the relationship of such behavior to more common phenomena, such as mixed valence, narrow-band effects, etc. In this paper we report bulk property measurements for a number of alloys which form a part of such a systematic study. The approach has been to take relatively simple and well-understood materials and alter their behavior by alloying to produce heavy fermion or Kondo behavior in a controlled way

  7. Thermodynamic properties of uranium--mercury system

    International Nuclear Information System (INIS)

    Lee, T.S.

    1979-01-01

    The EMF values in the fused salt cells of the type U(α)/KCl--LiCl--BaCl 2 eutectic, UCl 3 /U--Hg alloy, for the different two-phase alloys in the uranium--mercury system have been measured and the thermodynamic properties of this system have been calculated. These calculated values are in good agreement with values based on mercury vapor pressure measurements made by previous investigators. The inconsistency of the thermodynamic properties with the phase diagram determined by Frost are also confirmed. A tentative phase diagram based on the thermodynamic properties measured in this work was constructed

  8. Simple Magnetic Device Indicates Thickness Of Alloy 903

    Science.gov (United States)

    Long, Pin Jeng; Rodriguez, Sergio; Bright, Mark L.

    1995-01-01

    Handheld device called "ferrite indicator" orginally designed for use in determining ferrite content of specimen of steel. Placed in contact with specimen and functions by indicating whether magnet attracted more strongly to specimen or to calibrated reference sample. Relative strength of attraction shows whether alloy overlay thinner than allowable.

  9. Annex 5 - Fabrication of U-Al alloy

    International Nuclear Information System (INIS)

    Drobnjak, Dj.; Lazarevic, Dj.; Mihajlovic, A.

    1961-01-01

    Alloy U-Al with low content of aluminium is often used for fabrication of fuel elements because it is stable under moderate neutron flux density. Additionally this type of alloys show much better characteristics than pure uranium under reactor operating conditions (temperature, mechanical load, corrosion effect of water). This report contains the analysis of the phase diagram of U-Al alloy with low content of aluminium, applied procedure for alloying and casting with detailed description of equipment. Characteristics of the obtained alloy are described and conclusions about the experiment and procedure are presented [sr

  10. DNA extraction from herbarium specimens.

    Science.gov (United States)

    Drábková, Lenka Záveská

    2014-01-01

    With the expansion of molecular techniques, the historical collections have become widely used. Studying plant DNA using modern molecular techniques such as DNA sequencing plays an important role in understanding evolutionary relationships, identification through DNA barcoding, conservation status, and many other aspects of plant biology. Enormous herbarium collections are an important source of material especially for specimens from areas difficult to access or from taxa that are now extinct. The ability to utilize these specimens greatly enhances the research. However, the process of extracting DNA from herbarium specimens is often fraught with difficulty related to such variables as plant chemistry, drying method of the specimen, and chemical treatment of the specimen. Although many methods have been developed for extraction of DNA from herbarium specimens, the most frequently used are modified CTAB and DNeasy Plant Mini Kit protocols. Nine selected protocols in this chapter have been successfully used for high-quality DNA extraction from different kinds of plant herbarium tissues. These methods differ primarily with respect to their requirements for input material (from algae to vascular plants), type of the plant tissue (leaves with incrustations, sclerenchyma strands, mucilaginous tissues, needles, seeds), and further possible applications (PCR-based methods or microsatellites, AFLP).

  11. Uranium XAFS analysis of kidney from rats exposed to uranium.

    Science.gov (United States)

    Kitahara, Keisuke; Numako, Chiya; Terada, Yasuko; Nitta, Kiyohumi; Shimada, Yoshiya; Homma-Takeda, Shino

    2017-03-01

    The kidney is the critical target of uranium exposure because uranium accumulates in the proximal tubules and causes tubular damage, but the chemical nature of uranium in kidney, such as its chemical status in the toxic target site, is poorly understood. Micro-X-ray absorption fine-structure (µXAFS) analysis was used to examine renal thin sections of rats exposed to uranyl acetate. The U L III -edge X-ray absorption near-edge structure spectra of bulk renal specimens obtained at various toxicological phases were similar to that of uranyl acetate: their edge position did not shift compared with that of uranyl acetate (17.175 keV) although the peak widths for some kidney specimens were slightly narrowed. µXAFS measurements of spots of concentrated uranium in the micro-regions of the proximal tubules showed that the edge jump slightly shifted to lower energy. The results suggest that most uranium accumulated in kidney was uranium (VI) but a portion might have been biotransformed in rats exposed to uranyl acetate.

  12. Provision by the uranium and uranium products

    International Nuclear Information System (INIS)

    Elagin, Yu.P.

    2005-01-01

    International uranium market is converted from the buyer market into the seller market. The prices of uranium are high and the market attempts to adapt to changing circumstances. The industry of uranium enrichment satisfies the increasing demands but should to increase ots capacities. On the whole the situation is not stable and every year may change the existing position [ru

  13. Uranium recovery from slags of metallic uranium

    International Nuclear Information System (INIS)

    Fornarolo, F.; Frajndlich, E.U.C.; Durazzo, M.

    2006-01-01

    The Center of the Nuclear Fuel of the Institute of Nuclear Energy Research - IPEN finished the program of attainment of fuel development for research reactors the base of Uranium Scilicet (U 3 Si 2 ) from Hexafluoride of Uranium (UF 6 ) with enrichment 20% in weight of 235 U. In the process of attainment of the league of U 3 Si 2 we have as Uranium intermediate product the metallic one whose attainment generates a slag contend Uranium. The present work shows the results gotten in the process of recovery of Uranium in slags of calcined slags of Uranium metallic. Uranium the metallic one is unstable, pyrophoricity and extremely reactive, whereas the U 3 O 8 is a steady oxide of low chemical reactivity, what it justifies the process of calcination of slags of Uranium metallic. The calcination of the Uranium slag of the metallic one in oxygen presence reduces Uranium metallic the U 3 O 8 . Experiments had been developed varying it of acid for Uranium control and excess, nitric molar concentration gram with regard to the stoichiometric leaching reaction of temperature of the leaching process. The 96,0% income proves the viability of the recovery process of slags of Uranium metallic, adopting it previous calcination of these slags in nitric way with low acid concentration and low temperature of leaching. (author)

  14. Determination of uranium traces in fuel cans of nuclear reactors

    International Nuclear Information System (INIS)

    Acosta L, C.E.; Benavides M, A.M.; Sanchez P, L.A.; Nava S, G.F.

    1997-01-01

    The objective of this work is to quantify the uranium content that as impurity can be found in zircon and zircaloy alloys which are used in the construction of fuel cans. The determination of this serves as a quality control measure due to that the increment of uranium content in alloy, diminishing the corrosion resistance. The fluorimetric method was used to do this determination. It is a very sensitive, reliable, rapid method also high reproducibility and repeatability as well as low detection limits (0.25 mg/kg). (Author)

  15. Uranium enrichment

    International Nuclear Information System (INIS)

    Mohrhauer, H.

    1982-01-01

    The separation of uranium isotopes in order to enrich the fuel for light water reactors with the light isotope U-235 is an important part of the nuclear fuel cycle. After the basic principals of isotope separation the gaseous diffusion and the centrifuge process are explained. Both these techniques are employed on an industrial scale. In addition a short review is given on other enrichment techniques which have been demonstrated at least on a laboratory scale. After some remarks on the present situation on the enrichment market the progress in the development and the industrial exploitation of the gas centrifuge process by the trinational Urenco-Centec organisation is presented. (orig.)

  16. Effect of molybdenum addition on metastability of cubic γ-uranium

    International Nuclear Information System (INIS)

    Sinha, V.P.; Hegde, P.V.; Prasad, G.J.; Dey, G.K.; Kamath, H.S.

    2010-01-01

    Over the years U 3 Si 2 compound dispersed in aluminium matrix has been used successfully as the potential low enriched uranium (LEU 235 ) base dispersion fuel for use in new research and test reactors and also for converting high enriched uranium (HEU > 85%U 235 ) cores to LEU for most of the existing research and test reactors world over, though maximum 4.8 g U cm -3 density is achievable with U 3 Si 2 -Al dispersion fuel. To achieve a uranium density of 8.0-9.0 g U cm -3 in dispersion fuel with aluminium as matrix material, it is required to use γ-stabilized uranium metal powders. At Bhabha Atomic Research Centre (BARC), R and D efforts are on to develop these high density uranium base alloys. This paper describes the alloying behaviour of uranium with varying amount of molybdenum. The U-Mo alloys with different molybdenum content have been prepared by using an induction melting furnace with uranium and molybdenum metal pellets as starting materials. U-Mo alloys with different molybdenum content were characterized by X-ray diffraction (XRD) for phase identification and lattice parameter measurements. The optical microstructure of different U-Mo alloy composition has also been discussed in this paper. Quantitative image analysis was also carried out to determine the amount of various phases in each composition.

  17. Aqueous corrosion study on U-Zr alloy

    International Nuclear Information System (INIS)

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

    2009-01-01

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

  18. The hydrolysis of thorium dicarbide and of mixed uranium-thorium dicarbides

    International Nuclear Information System (INIS)

    Del Litto, B.

    1966-09-01

    The hydrolysis of thorium dicarbide leads to the formation of a complex mixture of gaseous and condensed carbon hydrides. The temperature, between 25 and 100 deg. C, has no influence on the nature and composition of the gas phase. The reaction kinetics, however, are strongly temperature dependent. In a hydrochloric medium, an enrichment in hydrogen of the gas mixture is observed. On the other hand a decrease in hydrogen and an increase in acetylene content take place in an oxidizing medium. The general results can be satisfactorily interpreted through a reaction mechanism involving C-C radical groups. In the same way, the hydrolysis of uranium-thorium-carbon ternary alloys leads to the formation of gaseous and condensed carbon hydrides. The variation of the composition of the gas phase versus uranium content in the alloy suggests an hypothesis about the carbon-carbon distance in the alloy crystal lattice. The variation of methane content, on the other hand, has lead us to discuss the nature of the various phases present in uranium-carbon alloys and carbon-rich uranium-thorium-carbon alloys. We have reached the conclusion that these alloys include a proportion of monocarbide which is dependent upon the ratio. Th/(Th + U). We put forward a diagram of the system uranium-carbon with features proper to explain some phenomena which have been observed in the uranium-thorium-carbon ternary diagram. (author) [fr

  19. Uranium conversion; Urankonvertering

    Energy Technology Data Exchange (ETDEWEB)

    Oliver, Lena; Peterson, Jenny; Wilhelmsen, Katarina [Swedish Defence Research Agency (FOI), Stockholm (Sweden)

    2006-03-15

    FOI, has performed a study on uranium conversion processes that are of importance in the production of different uranium compounds in the nuclear industry. The same conversion processes are of interest both when production of nuclear fuel and production of fissile material for nuclear weapons are considered. Countries that have nuclear weapons ambitions, with the intention to produce highly enriched uranium for weapons purposes, need some degree of uranium conversion capability depending on the uranium feed material available. This report describes the processes that are needed from uranium mining and milling to the different conversion processes for converting uranium ore concentrate to uranium hexafluoride. Uranium hexafluoride is the uranium compound used in most enrichment facilities. The processes needed to produce uranium dioxide for use in nuclear fuel and the processes needed to convert different uranium compounds to uranium metal - the form of uranium that is used in a nuclear weapon - are also presented. The production of uranium ore concentrate from uranium ore is included since uranium ore concentrate is the feed material required for a uranium conversion facility. Both the chemistry and principles or the different uranium conversion processes and the equipment needed in the processes are described. Since most of the equipment that is used in a uranium conversion facility is similar to that used in conventional chemical industry, it is difficult to determine if certain equipment is considered for uranium conversion or not. However, the chemical conversion processes where UF{sub 6} and UF{sub 4} are present require equipment that is made of corrosion resistant material.

  20. Issues in uranium availability

    International Nuclear Information System (INIS)

    Schanz, J.J. Jr.; Adams, S.S.; Gordon, R.L.

    1982-01-01

    The purpose of this publication is to show the process by which information about uranium reserves and resources is developed, evaluated and used. The following three papers in this volume have been abstracted and indexed for the Energy Data Base: (1) uranium reserve and resource assessment; (2) exploration for uranium in the United States; (3) nuclear power, the uranium industry, and resource development

  1. Australian uranium industry

    Energy Technology Data Exchange (ETDEWEB)

    Warner, R K

    1976-04-01

    Various aspects of the Australian uranium industry are discussed including the prospecting, exploration and mining of uranium ores, world supply and demand, the price of uranium and the nuclear fuel cycle. The market for uranium and the future development of the industry are described.

  2. Irradiated uranium reprocessing

    International Nuclear Information System (INIS)

    Gal, I.

    1961-12-01

    Task concerned with reprocessing of irradiated uranium covered the following activities: implementing the method and constructing the cell for uranium dissolving; implementing the procedure for extraction of uranium, plutonium and fission products from radioactive uranium solutions; studying the possibilities for using inorganic ion exchangers and adsorbers for separation of U, Pu and fission products

  3. Uranium processing and properties

    CERN Document Server

    2013-01-01

    Covers a broad spectrum of topics and applications that deal with uranium processing and the properties of uranium Offers extensive coverage of both new and established practices for dealing with uranium supplies in nuclear engineering Promotes the documentation of the state-of-the-art processing techniques utilized for uranium and other specialty metals

  4. Recovering uranium from phosphates

    Energy Technology Data Exchange (ETDEWEB)

    Bergeret, M [Compagnie de Produits Chimiques et Electrometallurgiques Pechiney-Ugine Kuhlmann, 75 - Paris (France)

    1981-06-01

    Processes for the recovery of the uranium contained in phosphates have today become competitive with traditional methods of working uranium sources. These new possibilities will make it possible to meet more rapidly any increases in the demand for uranium: it takes ten years to start working a new uranium deposit, but only two years to build a recovery plant.

  5. Uranium enrichment plans

    International Nuclear Information System (INIS)

    Gagne, R.W.; Thomas, D.C.

    1977-01-01

    The status of existing uranium enrichment contracts in the US is reviewed and expected natural uranium requirements for existing domestic uranium enrichment contracts are evaluated. Uncertainty in natural uranium requirements associated with requirements-type and fixed-commitment type contracts is discussed along with implementation of variable tails assay

  6. Uranium enrichment plans

    International Nuclear Information System (INIS)

    Thomas, D.C.; Gagne, R.W.

    1978-01-01

    The following topics are covered: the status of the Government's existing uranium enrichment services contracts, natural uranium requirements based on the latest contract information, uncertainty in predicting natural uranium requirements based on uranium enrichment contracts, and domestic and foreign demand assumed in enrichment planning

  7. Influence of thermal conditioning media on Charpy specimen test temperature

    International Nuclear Information System (INIS)

    Nanstad, R.K.; Swain, R.L.; Berggren, R.G.

    1989-01-01

    The Charpy V-notch (CVN) impact test is used extensively for determining the toughness of structural materials. Research programs in many technologies concerned with structural integrity perform such testing to obtain Charpy energy vs temperature curves. American Society for Testing and Materials Method E 23 includes rather strict requirements regarding determination and control of specimen test temperature. It specifies minimum soaking times dependent on the use of liquids or gases as the medium for thermally conditioning the specimen. The method also requires that impact of the specimen occur within 5 s removal from the conditioning medium. It does not, however, provide guidance regarding choice of conditioning media. This investigation was primarily conducted to investigate the changes in specimen temperature which occur when water is used for thermal conditioning. A standard CVN impact specimen of low-alloy steel was instrumented with surface-mounted and embedded thermocouples. Dependent on the media used, the specimen was heated or cooled to selected temperatures in the range -100 to 100 degree C using cold nitrogen gas, heated air, acetone and dry ice, methanol and dry ice, heated oil, or heated water. After temperature stabilization, the specimen was removed from the conditioning medium while the temperatures were recorded four times per second from all thermocouples using a data acquisition system and a computer. The results show that evaporative cooling causes significant changes in the specimen temperatures when water is used for conditioning. Conditioning in the other media did not result in such significant changes. The results demonstrate that, even within the guidelines of E 23, significant test temperature changes can occur which may substantially affect the Charpy impact test results if water is used for temperature conditioning. 7 refs., 11 figs

  8. Uranium industry annual 1985

    International Nuclear Information System (INIS)

    1986-11-01

    This report consists of two major sections. The first addresses uranium raw materials activities and covers the following topics: exploration activities and expenditures, resources and reserves, mine production of uranium, production of uranium concentrate, and industry employment. The second major section is concerned with the following uranium marketing activities: uranium purchase commitments, uranium prices, procurement arrangements, uranium imports and exports, enrichment services, inventories, secondary market activities, utility market requirements, and related topics. A glossary and appendices are included to assist the reader in interpreting the substantial array of statistical data in this report and to provide background information about the survey

  9. Uranium industry framework

    International Nuclear Information System (INIS)

    Riley, K.

    2008-01-01

    The global uranium market is undergoing a major expansion due to an increase in global demand for uranium, the highest uranium prices in the last 20 years and recognition of the potential greenhouse benefits of nuclear power. Australia holds approximately 27% of the world's uranium resources (recoverable at under US$80/kg U), so is well placed to benefit from the expansion in the global uranium market. Increasing exploration activity due to these factors is resulting in the discovery and delineation of further high grade uranium deposits and extending Australia's strategic position as a reliable and safe supplier of low cost uranium.

  10. Reduction of uranium hexafluoride to uranium tetrafluoride

    International Nuclear Information System (INIS)

    Chang, I.S.; Do, J.B.; Choi, Y.D.; Park, M.H.; Yun, H.H.; Kim, E.H.; Kim, Y.W.

    1982-01-01

    The single step continuous reduction of uranium hexafluoride (UF 6 ) to uranium tetrafluoride (UF 4 ) has been investigated. Heat required to initiate and maintain the reaction in the reactor is supplied by the highly exothermic reaction of hydrogen with a small amount of elemental fluorine which is added to the uranium hexafluoride stream. When gases uranium hexafluoride and hydrogen react in a vertical monel pipe reactor, the green product, UF 4 has 2.5g/cc in bulk density and is partly contaminated by incomplete reduction products (UF 5 ,U 2 F 9 ) and the corrosion product, presumably, of monel pipe of the reactor itself, but its assay (93% of UF 4 ) is acceptable for the preparation of uranium metal with magnesium metal. Remaining problems are the handling of uranium hexafluoride, which is easily clogging the flowmeter and gas feeding lines because of extreme sensitivity toward moisture, and a development of gas nozzel for free flow of uranium hexafluoride gas. (Author)

  11. Determination of uranium by a gravimetric-volumetric titration method

    International Nuclear Information System (INIS)

    Krtil, J.

    1998-01-01

    A volumetric-gravimetric modification of a method for the determination of uranium based on the reduction of uranium to U (IV) in a phosphoric acid medium and titration with a standard potassium dichromate solution is described. More than 99% of the stoichiometric amount of the titrating solution is weighed and the remainder is added volumetrically by using the Mettler DL 40 RC Memotitrator. Computer interconnected with analytical balances collects continually the data on the analyzed samples and evaluates the results of determination. The method allows to determine uranium in samples of uranium metal, alloys, oxides, and ammonium diuranate by using aliquot portions containing 30 - 100 mg of uranium with the error of determination, expressed as the relative standard deviation, of 0.02 - 0.05%. (author)

  12. Production of Mo-99 using low-enriched uranium silicide

    International Nuclear Information System (INIS)

    Hutter, J.C.; Srinivasan, B.; Vicek, M.; Vandegrift, G.F.

    1994-01-01

    Over the last several years, uranium silicide fuels have been under development as low-enriched uranium (LEU) targets for Mo-99. The use of LEU silicide is aimed at replacing the UAl x alloy in the highly-enriched uranium dissolution process. A process to recover Mo-99 from low-enriched uranium silicide is being developed at Argonne National Laboratory. The uranium silicide is dissolved in alkaline hydrogen peroxide. Experiments performed to determine the optimum dissolution procedure are discussed, and the results of dissolving a portion of a high-burnup (>40%) U 3 Si 2 miniplate are presented. Future work related to Mo-99 separation and waste disposal are also discussed

  13. Uranium - the world picture

    International Nuclear Information System (INIS)

    Silver, J.M.; Wright, W.J.

    1976-01-01

    The world resources of uranium and the future demand for uranium are discussed. The amount of uranium available depends on the price which users are prepared to pay for its recovery. As the price is increased, there is an incentive to recover uranium from lower grade or more difficult deposits. In view of this, attention is drawn to the development of the uranium industry in Australias

  14. An acceleration test for stress corrosion cracking using humped specimen

    International Nuclear Information System (INIS)

    Kamaya, Masayuki; Fukumura, Takuya; Totsuka, Nobuo

    2003-01-01

    By using the humped specimen, which is processed by the humped die, in the slow strain rate technique (SSRT) test, fracture facet due to stress corrosion cracking (SCC) can be observed in relatively short duration. Although the cold work and concentrated stress and strain caused by the characteristic shape of the specimen accelerate the SCC, to date these acceleration effects have not been examined quantitatively. In the present study, the acceleration effects of the humped specimen were examined through experiments and finite element analyses (FEA). The experiments investigated the SCC of alloy 600 in the primary water environment of a pressurized water reactor. SSRT tests were conducted using two kinds of humped specimen: one was annealed after hump processing in order to eliminate the cold work, and the other was hump processed after the annealing treatment. The work ratio caused by the hump processing and stress/strain conditions during SSRT test were evaluated by FEA. It was found that maximum work ratio of 30% is introduced by the hump processing and that the distribution of the work ratio is not uniform. Furthermore, the work ratio is influenced by the friction between the specimen and dies as well as by the shape of dies. It was revealed that not only the cold work but also the concentrated stress and strain during SSRT test accelerate the crack initiation and growth of the SCC. (author)

  15. Anticorrosion ion implantation of fragments of zirconium fuel can specimens

    International Nuclear Information System (INIS)

    Kalin, B.A.; Osipov, V.V.; Volkov, N.V.; Khernov, V.Yu.

    2001-01-01

    Aimed at the study of specific features of oxide film formation in the initial stage of Eh110 and Eh635 alloy fuel can oxidation the modification of tubular specimen surfaces is performed using an ion mixing technique, and the structure of oxide films produced in a steam-water environment is investigated. Using the method of vacuum vapor deposition the outer surface of specimens is coated with alloying element films irradiated by a polyenergetic Ar + ion beam with a 10 keV mean energy up to radiation doses of (7-10) x 10 17 ion/cm 2 . Monatomic (Al, Fe, Cu, Cr, Mo, Sn) or diatomic (Al-Fe, Al-Mo, Al-Sn, Fe-Cu, Fe-Mo, Fe-Sn, Cr-Mo, Cr-Sn) implantation into a zirconium cladding occurs under irradiation effect. The positive influence of combined intrusion of Al and other elements is revealed. The presence of Al atoms enhances the oxide film structure. The least ZeO 2 film thickness is observed when alloying with molybdenum, Al-Fe, Al-Mo and Al-Sn [ru

  16. Natural uranium

    International Nuclear Information System (INIS)

    Ammerich, Marc; Frot, Patricia; Gambini, Denis-Jean; Gauron, Christine; Moureaux, Patrick; Herbelet, Gilbert; Lahaye, Thierry; Pihet, Pascal; Rannou, Alain

    2014-08-01

    This sheet belongs to a collection which relates to the use of radionuclides essentially in unsealed sources. Its goal is to gather on a single document the most relevant information as well as the best prevention practices to be implemented. These sheets are made for the persons in charge of radiation protection: users, radioprotection-skill persons, labor physicians. Each sheet treats of: 1 - the radio-physical and biological properties; 2 - the main uses; 3 - the dosimetric parameters; 4 - the measurement; 5 - the protection means; 6 - the areas delimitation and monitoring; 7 - the personnel classification, training and monitoring; 8 - the effluents and wastes; 9 - the authorization and declaration administrative procedures; 10 - the transport; and 11 - the right conduct to adopt in case of incident or accident. This sheet deals specifically with natural uranium

  17. Irradiation-assisted stress corrosion cracking in HTH Alloy X-750 and Alloy 625

    International Nuclear Information System (INIS)

    Bajaj, R.; Mills, W.J.; Lebo, M.R.; Hyatt, B.Z.; Burke, M.G.

    1995-01-01

    In-reactor testing of bolt-loaded compact tension specimens was performed in 360 C water to determine the irradiation-assisted stress corrosion cracking (IASCC) behavior of HTH Alloy X-750 and direct-aged Alloy 625. New data confirm previous results showing that high irradiation levels reduce SCC resistance in Alloy X-750. Heat-to-heat variability correlates with boron content, with low boron heats showing improved IASCC properties. Alloy 625 is resistant to IASCC, as no cracking was observed in any Alloy 625 specimens. Microstructural, microchemical and deformation studies were performed to characterize the mechanisms responsible for IASCC in Alloy X-750 and the lack of an effect in Alloy 625. The mechanisms under investigation are: boron transmutation effects, radiation-induced changes in microstructure and deformation characteristics, and radiation-induced segregation. Irradiation of Alloy X-750 caused significant strengthening and ductility loss that was associated with the formation of cavities and dislocation loops. High irradiation levels did not cause significant segregation of alloying or trace elements in Alloy X-750. Irradiation of Alloy 625 resulted in the formation of small dislocation loops and a fine body-centered-orthorhombic phase. The strengthening due to the loops and precipitates was apparently offset by a partial dissolution of γ double-prime precipitates, as Alloy 625 showed no irradiation-induced strengthening or ductility loss. In the nonirradiated condition, an IASCC susceptible HTH heat containing 28 ppm B showed grain boundary segregation of boron, whereas a nonsusceptible HTH heat containing 2 ppm B and Alloy 625 with 20 ppm B did not show significant boron segregation. Transmutation of boron to helium at grain boundaries, coupled with matrix strengthening, is believed to be responsible for IASCC in Alloy X-750, and the absence of these two effects results in the superior IASCC resistance displayed by Alloy 625

  18. Effect of small additions of silicon, iron, and aluminum on the room-temperature tensile properties of high-purity uranium

    International Nuclear Information System (INIS)

    Ludwig, R.L.

    1983-01-01

    Eleven binary and ternary alloys of uranium and very low concentrations of iron, silicon, and aluminum were prepared and tested for room-temperature tensile properties after various heat treatments. A yield strength approximately double that of high-purity derby uranium was obtained from a U-400 ppM Si-200 ppM Fe alloy after beta solution treatment and alpha aging. Higher silicon plus iron alloy contents resulted in increased yield strength, but showed an unacceptable loss of ductility

  19. Variation of the uranium monocarbide parameter with changes in the carbon content; Variations du parametre du monocarbure d'uranium en fonction de sa teneur en carbone

    Energy Technology Data Exchange (ETDEWEB)

    Magnier, P; Accary, A [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1965-07-01

    The authors show that the chemical species uranium monocarbide is only a particular composition of the uranium-carbon alloy phase containing between 48 and 50 atoms per cent of carbon, and that the crystalline parameter of this phase varies simultaneously from 4.956 to 4.961 Angstroms. (authors) [French] Les auteurs montrent que l'espece chimique monocarbure d'uranium n'est qu'une composition particuliere de la phase des alliages uranium carbone contenant entre 48 et 50 atomes pour cent de carbone et que le parametre cristallin de cette phase varie simultanement de 4.956 a 4.961 Angstroms.

  20. Tensile tests and metallography of brazed AISI 316L specimens after irradiation

    International Nuclear Information System (INIS)

    Groot, P.; Franconi, E.

    1994-01-01

    Stainless steel type 316L tensile specimens were vacuum brazed with three kinds of alloys: BNi-5, BNi-6, and BNi-7. The specimens were irradiated up to 0.7 dpa at 353 K in the High Flux Reactor at JRC Petten, the Netherlands. Tensile tests were performed at a constant displacement rate of 10 -3 s -1 at room temperature in the ECN hot cell facility. BNi-5 brazed specimens showed ductile behaviour. Necking and fractures were localized in the plate material. BNi-6 and BNi-7 brazed specimens failed brittle in the brazed zone. This was preceded by uniform deformation of the plate material. Tensile test results of irradiated specimens showed higher stresses due to radiation hardening and a reduction of the elongation of the plate material compared to the reference. SEM examination of the irradiated BNi-6 and BNi-7 fracture surfaces showed nonmetallic phases. These phases were not found in the reference specimens. ((orig.))

  1. ASTM international symposium on small specimen test techniques and their applications to pressure vessel annealing and plant life extension

    International Nuclear Information System (INIS)

    Garner, F.A.; Hamilton, M.L.; Heinisch, H.L.; Kumar, A.S.

    1992-01-01

    Miniature sheet-type tensile specimens are currently being used in a variety of radiation damage studies conducted in a number of different reactors. Although these specimens are very small, they have proven successful in addressing issues encountered in both thermal reactors and anticipated fusion reactors. This paper reviews the results of a number of recent studies that illustrate the range of applicability of these small specimens. When combined with other types of specimens and other types of measurements made prior to tensile testing, miniature tensile specimens have been found to serve as very useful tools for application to both fundamental studies and alloy screening studies

  2. Summary of the U.S. specimen matrix for the HFIR 13J varying temperature irradiation capsule

    International Nuclear Information System (INIS)

    Zinkle, S.J.

    1998-01-01

    The US specimen matrix for the collaborative DOE/Monbusho HFIR 13J varying temperature irradiation capsule contains two ceramics and 29 different metals, including vanadium alloys, ferritic/martensitic steels, pure iron, austenitic stainless steels, nickel alloys, and copper alloys. This experiment is designed to provide fundamental information on the effects of brief low-temperature excursions on the tensile properties and microstructural evolution of a wide range of materials irradiated at nominal temperatures of 350 and 500 C to a dose of ∼5 dpa. A total of 340 miniature sheet tensile specimens and 274 TEM disks are included in the US-supplied matrix for the irradiation capsule

  3. Simple process to fabricate nitride alloy powders

    International Nuclear Information System (INIS)

    Yang, Jae Ho; Kim, Dong-Joo; Kim, Keon Sik; Rhee, Young Woo; Oh, Jang-Soo; Kim, Jong Hun; Koo, Yang Hyun

    2013-01-01

    Uranium mono-nitride (UN) is considered as a fuel material [1] for accident-tolerant fuel to compensate for the loss of fissile fuel material caused by adopting a thickened cladding such as SiC composites. Uranium nitride powders can be fabricated by a carbothermic reduction of the oxide powders, or the nitriding of metal uranium. Among them, a direct nitriding process of metal is more attractive because it has advantages in the mass production of high-purity powders and the reusing of expensive 15 N 2 gas. However, since metal uranium is usually fabricated in the form of bulk ingots, it has a drawback in the fabrication of fine powders. The Korea Atomic Energy Research Institute (KAERI) has a centrifugal atomisation technique to fabricate uranium and uranium alloy powders. In this study, a simple reaction method was tested to fabricate nitride fuel powders directly from uranium metal alloy powders. Spherical powder and flake of uranium metal alloys were fabricated using a centrifugal atomisation method. The nitride powders were obtained by thermal treating the metal particles under nitrogen containing gas. The phase and morphology evolutions of powders were investigated during the nitriding process. A phase analysis of nitride powders was also part of the present work. KAERI has developed the centrifugal rotating disk atomisation process to fabricate spherical uranium metal alloy powders which are used as advanced fuel materials for research reactors. The rotating disk atomisation system involves the tasks of melting, atomising, and collecting. A nozzle in the bottom of melting crucible introduces melt at the center of a spinning disk. The centrifugal force carries the melt to the edge of the disk and throws the melt off the edge. Size and shape of droplets can be controlled by changing the nozzle size, the disk diameter and disk speed independently or simultaneously. By adjusting the processing parameters of the centrifugal atomiser, a spherical and flake shape

  4. Computer simulation of structural modifications induced by highly energetic ions in uranium dioxide

    Energy Technology Data Exchange (ETDEWEB)

    Sasajima, Y., E-mail: sasajima@mx.ibaraki.ac.jp [Department of Materials Science and Engineering, Faculty of Engineering, Ibaraki University, 4-12-1 Nakanarusawa, Hitachi 316-8511 (Japan); Frontier Research Center for Applied Atomic Sciences, Ibaraki University, Shirakata 162-4, Tokai 319-1106 (Japan); Osada, T. [Graduate School of Science and Engineering, Ibaraki University, 4-12-1 Nakanarusawa, Hitachi 316-8511 (Japan); Ishikawa, N. [Japan Atomic Energy Agency (JAEA), Shirakata Shirane 2-4, Tokai 319-1195 (Japan); Iwase, A. [Department of Materials Science, Osaka Prefecture University, Gakuen-cho 1-1, Sakai 599-8531 (Japan)

    2013-11-01

    The structural modification caused by the high-energy-ion irradiation of single-crystalline uranium dioxide was simulated by the molecular dynamics method. As the initial condition, high kinetic energy was supplied to the individual atoms within a cylindrical region of nanometer-order radius located in the center of the specimen. The potential proposed by Basak et al. [C.B. Basak, A.K. Sengupta, H.S. Kamath, J. Alloys Compd. 360 (2003) 210–216] was utilized to calculate interaction between atoms. The supplied kinetic energy was first spent to change the crystal structure into an amorphous one within a short period of about 0.3 ps, then it dissipated in the specimen. The amorphous track radius R{sub a} was determined as a function of the effective stopping power gS{sub e}, i.e., the kinetic energy of atoms per unit length created by ion irradiation (S{sub e}: electronic stopping power, g: energy transfer ratio from stopping power to lattice vibration energy). It was found that the relationship between R{sub a} and gS{sub e} follows the relation R{sub a}{sup 2}=aln(gS{sub e})+b. Compared to the case of Si and β-cristobalite single crystals, it was harder to produce amorphous track because of the long range interaction between U atoms.

  5. Computer simulation of structural modifications induced by highly energetic ions in uranium dioxide

    International Nuclear Information System (INIS)

    Sasajima, Y.; Osada, T.; Ishikawa, N.; Iwase, A.

    2013-01-01

    The structural modification caused by the high-energy-ion irradiation of single-crystalline uranium dioxide was simulated by the molecular dynamics method. As the initial condition, high kinetic energy was supplied to the individual atoms within a cylindrical region of nanometer-order radius located in the center of the specimen. The potential proposed by Basak et al. [C.B. Basak, A.K. Sengupta, H.S. Kamath, J. Alloys Compd. 360 (2003) 210–216] was utilized to calculate interaction between atoms. The supplied kinetic energy was first spent to change the crystal structure into an amorphous one within a short period of about 0.3 ps, then it dissipated in the specimen. The amorphous track radius R a was determined as a function of the effective stopping power gS e , i.e., the kinetic energy of atoms per unit length created by ion irradiation (S e : electronic stopping power, g: energy transfer ratio from stopping power to lattice vibration energy). It was found that the relationship between R a and gS e follows the relation R a 2 =aln(gS e )+b. Compared to the case of Si and β-cristobalite single crystals, it was harder to produce amorphous track because of the long range interaction between U atoms

  6. Postirradiation notch ductility tests of ESR alloy HT-9 and modified 9Cr-1Mo alloy from UBR reactor experiments

    International Nuclear Information System (INIS)

    Hawthorne, J.R.

    1984-01-01

    During this period, irradiation exposures at 300 0 C and 150 0 C to approx. 8 x 10 19 n/cm 2 , E > 0.1 MeV, were completed for the Alloy HT-9 plate and the modified Alloy 9Cr-1Mo plates, respectively. Postirradiation tests of Charpy-V (C/sub v/) specimens were completed for both alloys; other specimen types included in the reactor assemblies were fatigue precracked Charpy-V (PCC/sub v/), half-size Charpy-V, and in the case of the modified 9Cr-1Mo, 2.54 mm thick compact tension specimens

  7. Uranium management activities

    International Nuclear Information System (INIS)

    Jackson, D.; Marshall, E.; Sideris, T.; Vasa-Sideris, S.

    2001-01-01

    One of the missions of the Department of Energy's (DOE) Oak Ridge Office (ORO) has been the management of the Department's uranium materials. This mission has been accomplished through successful integration of ORO's uranium activities with the rest of the DOE complex. Beginning in the 1980's, several of the facilities in that complex have been shut down and are in the decommissioning process. With the end of the Cold War, the shutdown of many other facilities is planned. As a result, inventories of uranium need to be removed from the Department facilities. These inventories include highly enriched uranium (HEU), low enriched uranium (LEU), normal uranium (NU), and depleted uranium (DU). The uranium materials exist in different chemical forms, including metals, oxides, solutions, and gases. Much of the uranium in these inventories is not needed to support national priorities and programs. (author)

  8. Uranium industry annual 1998

    International Nuclear Information System (INIS)

    1999-01-01

    The Uranium Industry Annual 1998 (UIA 1998) provides current statistical data on the US uranium industry's activities relating to uranium raw materials and uranium marketing. It contains data for the period 1989 through 2008 as collected on the Form EIA-858, ''Uranium Industry Annual Survey.'' Data provides a comprehensive statistical characterization of the industry's activities for the survey year and also include some information about industry's plans and commitments for the near-term future. Data on uranium raw materials activities for 1989 through 1998, including exploration activities and expenditures, EIA-estimated reserves, mine production of uranium, production of uranium concentrate, and industry employment, are presented in Chapter 1. Data on uranium marketing activities for 1994 through 2008, including purchases of uranium and enrichment services, enrichment feed deliveries, uranium fuel assemblies, filled and unfilled market requirements, and uranium inventories, are shown in Chapter 2. The methodology used in the 1998 survey, including data edit and analysis, is described in Appendix A. The methodologies for estimation of resources and reserves are described in Appendix B. A list of respondents to the ''Uranium Industry Annual Survey'' is provided in Appendix C. The Form EIA-858 ''Uranium Industry Annual Survey'' is shown in Appendix D. For the readers convenience, metric versions of selected tables from Chapters 1 and 2 are presented in Appendix E along with the standard conversion factors used. A glossary of technical terms is at the end of the report. 24 figs., 56 tabs

  9. Uranium industry annual 1994

    International Nuclear Information System (INIS)

    1995-01-01

    The Uranium Industry Annual 1994 (UIA 1994) provides current statistical data on the US uranium industry's activities relating to uranium raw materials and uranium marketing during that survey year. The UIA 1994 is prepared for use by the Congress, Federal and State agencies, the uranium and nuclear electric utility industries, and the public. It contains data for the 10-year period 1985 through 1994 as collected on the Form EIA-858, ''Uranium Industry Annual Survey.'' Data collected on the ''Uranium Industry Annual Survey'' (UIAS) provide a comprehensive statistical characterization of the industry's activities for the survey year and also include some information about industry's plans and commitments for the near-term future. Where aggregate data are presented in the UIA 1994, care has been taken to protect the confidentiality of company-specific information while still conveying accurate and complete statistical data. A feature article, ''Comparison of Uranium Mill Tailings Reclamation in the United States and Canada,'' is included in the UIA 1994. Data on uranium raw materials activities including exploration activities and expenditures, EIA-estimated resources and reserves, mine production of uranium, production of uranium concentrate, and industry employment are presented in Chapter 1. Data on uranium marketing activities, including purchases of uranium and enrichment services, and uranium inventories, enrichment feed deliveries (actual and projected), and unfilled market requirements are shown in Chapter 2

  10. Evolution of the uranium local environment during alteration of SON68 glass

    International Nuclear Information System (INIS)

    Jollivet, Patrick; Auwer, Christophe Den; Simoni, Eric

    2002-01-01

    The speciation of uranium in SON68 glass specimens doped with 0.75-3.5 wt% uranium and in the gels formed by alteration of the specimens was investigated by X-ray absorption spectroscopy. In the glasses, uranium is present at oxidation state VI and coordination number 6 with the same average distances than those found in a UO 3 type environment. The U-O distances and uranium coordination numbers are identical throughout the uranium concentration range. During glass alteration the uranium remains at oxidation state VI in the gels, but was found in the uranyl form. An increase in the equatorial distances (from 2.20 and 2.32 A in the glass to respectively 2.22 and 2.39 A in the gel) and coordination numbers (to about 7 and 8, respectively) was observed

  11. Fracture resistance of welded panel specimen with perpendicular crack in tensile

    International Nuclear Information System (INIS)

    Gochev, Todor; Adziev, Todor

    1998-01-01

    Defects caused by natural crack in welded joints of high-strength low-alloy (HSLA) steels are very often. Perpendicular crack in welded joints and its heat treatment after the welding has also an influence on the fracture resistance. The fracture resistance of welded joints by crack in tense panel specimens was investigated by crack mouse opening displesment (CMOD), the parameter of fracture mechanic. Crack propagation was analysed by using a metallographic analysis of fractured specimens after the test. (Author)

  12. Determination of the DBTT of Aluminide Coatings and its Influence on the Mechanical Behavior of Coated Specimens

    Energy Technology Data Exchange (ETDEWEB)

    Dryepondt, Sebastien N [ORNL; Pint, Bruce A [ORNL

    2010-01-01

    The ductility of various coatings deposited by chemical vapor deposition, pack cementation and slurry processes on Fe- and Ni-based alloys was characterized by indentation at room temperature. A hot indentation apparatus has also been developed to more rapidly determine the ductile to brittle transition temperature of coated specimens. Creep testing has been conducted on bare and coated alloy 230 (NiCrW) specimens at 800 C with a significant decrease in creep life observed. Based on the observed failure of coated 230 specimens, the impact of coating ductility on substrate creep properties is discussed.

  13. Uranium: a basic evaluation

    International Nuclear Information System (INIS)

    Crull, A.W.

    1978-01-01

    All energy sources and technologies, including uranium and the nuclear industry, are needed to provide power. Public misunderstanding of the nature of uranium and how it works as a fuel may jeopardize nuclear energy as a major option. Basic chemical facts about uranium ore and uranium fuel technology are presented. Some of the major policy decisions that must be made include the enrichment, stockpiling, and pricing of uranium. Investigations and lawsuits pertaining to uranium markets are reviewed, and the point is made that oil companies will probably have to divest their non-oil energy activities. Recommendations for nuclear policies that have been made by the General Accounting Office are discussed briefly

  14. Uranium health physics

    International Nuclear Information System (INIS)

    1980-01-01

    This report contains the papers delivered at the Summer School on Uranium Health Physics held in Pretoria on the 14 and 15 April 1980. The following topics were discussed: uranium producton in South Africa; radiation physics; internal dosimetry and radiotoxicity of long-lived uranium isotopes; uranium monitoring; operational experience on uranium monitoring; dosimetry and radiotoxicity of inhaled radon daughters; occupational limits for inhalation of radon-222, radon-220 and their short-lived daughters; radon monitoring techniques; radon daughter dosimeters; operational experience on radon monitoring; and uranium mill tailings management

  15. Uranium: one utility's outlook

    International Nuclear Information System (INIS)

    Gass, C.B.

    1983-01-01

    The perspective of the Arizona Public Service Company (APS) on the uncertainty of uranium as a fuel supply is discussed. After summarizing the history of nuclear power and the uranium industries, a projection is made for the future uranium market. An uncrtain uranium market is attributed to various determining factors that include international politics, production costs, non-commercial government regulation, production-company stability, and questionable levels of uranium sales. APS offers its solutions regarding type of contract, choice of uranium producers, pricing mechanisms, and aids to the industry as a whole. 5 references, 10 figures, 1 table

  16. Atomic-scale Studies of Uranium Oxidation and Corrosion by Water Vapour

    OpenAIRE

    T. L. Martin; C. Coe; P. A. J. Bagot; P. Morrall; G. D. W Smith; T. Scott; M. P. Moody

    2016-01-01

    Understanding the corrosion of uranium is important for its safe, long-term storage. Uranium metal corrodes rapidly in air, but the exact mechanism remains subject to debate. Atom Probe Tomography was used to investigate the surface microstructure of metallic depleted uranium specimens following polishing and exposure to moist air. A complex, corrugated metal-oxide interface was observed, with approximately 60 at.% oxygen content within the oxide. Interestingly, a very thin (∼5 nm) interfacia...

  17. The life of some metallic uranium based fuel elements; Duree de vie de quelques combustibles a base d'uranium metal

    Energy Technology Data Exchange (ETDEWEB)

    Stohr, J A; Englander, M [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1958-07-01

    Description of some theoretical and experimental data concerning the design and most economic preparation of metallic uranium based fuel elements, which are intended to produce an energy of 3 kW days/g of uranium in a thermal reactor, at a sufficiently high mean temperature. Experimental results obtained by testing by analogy or by actually trying out fuel elements obtained by alloying uranium with other metals in proportions such that the resistance to deformation of the alloy produced is much higher than that of pure metallic uranium and that the thermal utilisation factor is only slightly different from that of the uranium. (author) [French] Description de quelques donnees theoriques et experimentales concernant la conception et la preparation la plus economique d'elements combustibles a base d'uranium metallique naturel, destines a degager dans un reacteur thermique une energie de l'ordre de 3 kWj/g d'uranium a une temperature moyenne suffisamment elevee. Resultats experimentaux acquis par tests analogiques ou reels sur combustibles obtenus par alliage de l'uranium avec des elements metalliques en proportions telles que la resistance a la deformation soit bien superieure a celle de l'uranium metal pur et que le facteur propre d'utilisation thermique n ne soit que peu affecte. (auteur)

  18. Deformation of a Low-Cost Ti-6A1-4V Armor Alloy Under Shock Loading

    National Research Council Canada - National Science Library

    Spletzer, Stephen

    2001-01-01

    .... Examination of the particle velocity histories obtained from specimens of the alloy during 11 plate-on-plate impact/planar shock wave experiments indicates that the alloy deforms in an elastic-plastic manner...

  19. DUCTILE URANIUM FUEL FOR NUCLEAR REACTORS AND METHOD OF MAKING

    Science.gov (United States)

    Zegler, S.T.

    1963-11-01

    The fabrication process for a ductile nuclear fuel alloy consisting of uranium, fissium, and from 0.25 to 1.0 wt% of silicon or aluminum or from 0.25 to 2 wt% of titanium or yttrium is presented. (AEC)

  20. Recovery of uranium from crude uranium tetrafluoride

    International Nuclear Information System (INIS)

    Ghosh, S.K.; Bellary, M.P.; Keni, V.S.

    1994-01-01

    An innovative process has been developed for recovery of uranium from crude uranium tetrafluoride cake. The process is based on direct dissolution of uranium tetrafluoride in nitric acid in presence of aluminium hydroxide and use of solvent extraction for removal of fluorides and other bulk impurities to make uranium amenable for refining. It is a simple process requiring minimum process step and has advantage of lesser plant corrosion. This process can be applied for processing of uranium tetrafluoride generated from various sources like uranium by-product during thorium recovery from thorium concentrate, first stage product of uranium recovery from phosphoric acid by OPPA process and off grade uranium tetrafluoride material. The paper describes the details of the process developed and demonstrated on bench and pilot scale and its subsequent modification arising out of bulky solid waste generation. The modified process uses a lower quantity of aluminium hydroxide by allowing a lower dissolution of uranium per cycle and recycles the undissolved material to the next cycle, maintaining the overall recovery at high level. This innovation has reduced the solid waste generated by a factor of four at the cost of a slightly larger dissolution vessel and its increased corrosion rate. (author)

  1. Recovery of uranium from crude uranium tetrafluoride

    Energy Technology Data Exchange (ETDEWEB)

    Ghosh, S K; Bellary, M P; Keni, V S [Chemical Engineering Division, Bhabha Atomic Research Centre, Mumbai (India)

    1994-06-01

    An innovative process has been developed for recovery of uranium from crude uranium tetrafluoride cake. The process is based on direct dissolution of uranium tetrafluoride in nitric acid in presence of aluminium hydroxide and use of solvent extraction for removal of fluorides and other bulk impurities to make uranium amenable for refining. It is a simple process requiring minimum process step and has advantage of lesser plant corrosion. This process can be applied for processing of uranium tetrafluoride generated from various sources like uranium by-product during thorium recovery from thorium concentrate, first stage product of uranium recovery from phosphoric acid by OPPA process and off grade uranium tetrafluoride material. The paper describes the details of the process developed and demonstrated on bench and pilot scale and its subsequent modification arising out of bulky solid waste generation. The modified process uses a lower quantity of aluminium hydroxide by allowing a lower dissolution of uranium per cycle and recycles the undissolved material to the next cycle, maintaining the overall recovery at high level. This innovation has reduced the solid waste generated by a factor of four at the cost of a slightly larger dissolution vessel and its increased corrosion rate. (author). 4 refs., 1 fig., 3 tabs.

  2. Investigation of phase transformations of U2.5Zr7.5Nb and U3Zr9Nb alloys aging at 600 deg C

    International Nuclear Information System (INIS)

    Cantagalli, Natalia Mattar; Tanure, Leandro Paulo de Almeida Reis; Braga, Daniel Martins; Santos, Ana Maria Matildes dos; Ferraz, Wilmar Barbosa

    2009-01-01

    Investigation has been made of the effects of high-temperature aging (600 deg C) on the phase transformations in the U2.5Zr7.5Nb and U3Zr9Nb alloys. These alloys have been produced with vacuum induction melting (VIM) furnace in cast ingots. The ingots were homogenized at 1000 deg C for 24 hours in vacuum of -4 torr, and cooled to room temperature at a rate of 3 deg C/min. Specimens from these homogeneous materials, cut in 3 mm high and 10 mm diameter, were reheated to γ phase at 850 deg C, for 1 hour, and aging at 600 deg C at different times from 0.5 to 24 hours. The phases decomposition were characterized by X-ray diffraction (XRD), metallographic, micro-probe analyze by energy dispersive spectrometry (EDS) and microhardness methods. It was verified that the decomposition of the δ phase proceeds in two steps. The first is a discontinuous precipitation of a lamellar two-phase aggregate composed of alpha solid solution and a metastable gamma phase. The metastable gamma phase has a constant composition at given temperature. After longer annealing, it decomposes eutectoidally into the equilibrium (α + δ 2 ) phases mixture. During this process a modification of the original lamellar microstructure takes place. The obtained metastable phases of these alloys of different compositions were analyzed in relation to their constitution, heat treatability and micrographic features and the results confronted with available distinct uranium alloys data from literature. (author)

  3. Irradiation assisted stress corrosion cracking of HTH Alloy X-750 and Alloy 625

    International Nuclear Information System (INIS)

    Mills, W.J.; Lebo, M.R.; Bajaj, R.; Kearns, J.J.; Hoffman, R.C.; Korinko, J.J.

    1994-01-01

    In-reactor testing of bolt-loaded precracked compact tension specimens was performed in 360 degree C water to determine effect of irradiation on the SCC behavior of HTH Alloy X-750 and direct aged Alloy 625. Out-of-flux and autoclave control specimens provided baseline data. Primary test variables were stress intensity factor, fluence, chemistry, processing history, prestrain. Results for the first series of experiments were presented at a previous conference. Data from two more recent experiments are compared with previous results; they confirm that high irradiation levels significantly reduce SCC resistance in HTH Alloy X-750. Heat-to-heat differences in IASCC were related to differences in boron content, with low boron heats showing improved SCC resistance. The in-reactor SCC performance of Alloy 625 was superior to that for Alloy X-750, as no cracking was observed in any Alloy 625 specimens even though they were tested at very high K 1 and fluence levels. A preliminary SCC usage model developed for Alloy X-750 indicates that in-reactor creep processes, which relax stresses but also increase crack tip strain rates, and radiolysis effects accelerate SCC. Hence, in-reactor SCC damage under high flux conditions may be more severe than that associated with postirradiation tests. In addition, preliminary mechanism studies were performed to determine the cause of IASCC In Alloy X-750

  4. Uranium production

    International Nuclear Information System (INIS)

    Jones, J.Q.

    1981-01-01

    The domestic uranium industry is in a state of stagflation. Costs continue to rise while the market for the product remains stagnant. During the last 12 months, curtailments and closures of mines and mills have eliminated over 5000 jobs in the industry, plus many more in those industries that furnish supplies and services. By January 1982, operations at four mills and the mines that furnish them ore will have been terminated. Other closures may follow, depending on cost trends, duration of current contracts, the degree to which mills have been amortized, the feasibility of placing mines on standby, the grade of the ore, and many other factors. Open-pit mines can be placed on standby without much difficulty, other than the possible cost of restoration before all the ore has been removed. There are a few small, dry, underground mines that could be mothballed; however, the major underground producers are wet sandstone mines that in most cases could not be reopened after a prolonged shutdown; mills can be mothballed for several years. Figure 8 shows the location of all the production centers in operation, as well as those that have operated or are on standby. Table 1 lists the same production centers plus those that have been deferred, showing nominal capacity of conventional mills in tons of ore per calendar day, and the industry production rate for those mills as of October 1, 1981

  5. Thermal Conductivity of Metallic Uranium

    Energy Technology Data Exchange (ETDEWEB)

    Hin, Celine

    2018-03-10

    used in the original fitting. Moreover, as fuels burn up in the reactor and fission products are built up, thermal conductivity is also significantly changed [3]. Unfortunately, fundamental understanding of the effect of fission products is also currently lacking. In this project, we probe thermal conductivity of metallic fuels with ab initio calculations, a theoretical tool with the potential to yield better accuracy and predictive power than empirical fitting. This work will both complement experimental data by determining thermal conductivity in wider composition and temperature ranges than is available experimentally, and also develop mechanistic understanding to guide better design of metallic fuels in the future. So far, we focused on α-U perfect crystal, the ground-state phase of U metal. We focus on two methods. The first method has been developed by the team at the University of Wisconsin Madison. They developed a practical and general modeling approach for thermal conductivity of metals and metal alloys that integrates ab-initio and semi-empirical physics-based models to maximize the strengths of both techniques. The second method has been developed by the team at Virginia Tech. This approach consists of a determining the thermal conductivity using only ab-initio methods without any fitting parameters. Both methods were complementary and very helpful to understand the physics behind the thermal conductivity in metallic uranium and other materials with similar characteristics. In Section I, the combined model developed at UWM is explained. In Section II, the ab-initio method developed at VT is described along with the uranium pseudo-potential and its validation. Section III is devoted to the work done by Jianguo Yu at INL. Finally, we will present the performance of the project in terms of milestones, publications, and presentations.

  6. Change of Composition in Metallic Fuel Slug of U-Zr Alloy from High-Temperature Annealing

    Energy Technology Data Exchange (ETDEWEB)

    Youn, Young Sang; Lee, Jeong Mook; Kim, Jong Yun; Kim, Jong Hwan; Song, Hoon [KAERI, Daejeon (Korea, Republic of)

    2016-09-15

    The U–Zr alloy is a candidate for fuel to be used as metallic fuel in sodium-cooled fast reactors (SFRs). Its chemical composition before and after annealing at the operational temperature of SFRs (610 .deg. C) was investigated using X-ray photoelectron spectroscopy, Raman spectroscopy, and X-ray diffraction. The original alloy surface contained uranium oxides with the U(IV) and U(VI) oxidation states, Zr{sub 2}O{sub 3}, and a low amount of uranium metal. After annealing at 610 .deg. C, the alloy was composed of uranium metal, uranium carbide, uranium oxide with the U(V) valence state, zirconium metal, and amorphous carbon. Meanwhile, X-ray diffraction data indicate that the bulk composition of the alloy remained unchanged.

  7. Change of Composition in Metallic Fuel Slug of U-Zr Alloy from High-Temperature Annealing

    International Nuclear Information System (INIS)

    Youn, Young Sang; Lee, Jeong Mook; Kim, Jong Yun; Kim, Jong Hwan; Song, Hoon

    2016-01-01

    The U–Zr alloy is a candidate for fuel to be used as metallic fuel in sodium-cooled fast reactors (SFRs). Its chemical composition before and after annealing at the operational temperature of SFRs (610 .deg. C) was investigated using X-ray photoelectron spectroscopy, Raman spectroscopy, and X-ray diffraction. The original alloy surface contained uranium oxides with the U(IV) and U(VI) oxidation states, Zr 2 O 3 , and a low amount of uranium metal. After annealing at 610 .deg. C, the alloy was composed of uranium metal, uranium carbide, uranium oxide with the U(V) valence state, zirconium metal, and amorphous carbon. Meanwhile, X-ray diffraction data indicate that the bulk composition of the alloy remained unchanged

  8. Uranium mining in Australia

    International Nuclear Information System (INIS)

    Anon.

    1980-01-01

    Known uranium deposits and the companies involved in uranium mining and exploration in Australia are listed. The status of the development of the deposits is outlined and reasons for delays to mining are given

  9. Uranium Processing Facility

    Data.gov (United States)

    Federal Laboratory Consortium — An integral part of Y‑12's transformation efforts and a key component of the National Nuclear Security Administration's Uranium Center of Excellence, the Uranium...

  10. Uranium in Niger

    International Nuclear Information System (INIS)

    Gabelmann, E.

    1978-03-01

    This document presents government policy in the enhancement of uranium resources, existing mining companies and their productions, exploitation projects and economical outcome related to the uranium mining and auxiliary activities [fr

  11. Price of military uranium

    International Nuclear Information System (INIS)

    Klimenko, A.V.

    1998-01-01

    The theoretical results about optimum strategy of use of military uranium confirmed by systems approach accounts are received. The numerical value of the system approach price of the highly enriched military uranium also is given

  12. Uranium market and resources

    International Nuclear Information System (INIS)

    Capus, G.; Arnold, Th.

    2004-01-01

    The controversy about the extend of the uranium resources worldwide is still important, this article sheds some light on this topic. Every 2 years IAEA and NEA (nuclear energy agency) edit an inventory of uranium resources as reported by contributing countries. It appears that about 4.6 millions tons of uranium are available at a recovery cost less than 130 dollars per kg of uranium and a total of 14 millions tons of uranium can be assessed when including all existing or supposed resources. In fact there is enough uranium to sustain a moderate growth of the park of nuclear reactors during next decades and it is highly likely that the volume of uranium resources can allow a more aggressive development of nuclear energy. It is recalled that a broad use of the validated breeder technology can stretch the durability of uranium resources by a factor 50. (A.C.)

  13. Uranium from phosphate ores

    International Nuclear Information System (INIS)

    Hurst, F.J.

    1983-01-01

    The following topics are described briefly: the way phosphate fertilizers are made; how uranium is recovered in the phosphate industry; and how to detect covert uranium recovery operations in a phsophate plant

  14. Industrial realities: Uranium

    International Nuclear Information System (INIS)

    Thiron, H.

    1990-01-01

    In this special issue are examined ores and metals in France and in the world for 1988. The chapter on uranium gives statistical data on the uranium market: Demand, production, prices and reserves [fr

  15. Brazilian uranium deposits

    International Nuclear Information System (INIS)

    Santos, L.C.S. dos.

    1985-01-01

    Estimatives of uranium reserves carried out in Figueira, Itataia, Lagoa Real and Espinharas, in Brazil are presented. The samples testing allowed to know geological structures, and the characteristics of uranium mineralization. (M.C.F.) [pt

  16. Uranium mining in Australia

    International Nuclear Information System (INIS)

    Anon.

    1984-01-01

    The mining of uranium in Australia is criticised in relation to it's environmental impact, economics and effects on mine workers and Aborigines. A brief report is given on each of the operating and proposed uranium mines in Australia

  17. Polarographic methods for the analysis of beryllium metal and its alloys

    International Nuclear Information System (INIS)

    Wells, J.M.

    1975-10-01

    This report describes polarographic methods for the analysis of beryllium metal and its alloys. The elements covered by these methods are aluminium, bismuth, cadmium, cobalt, copper, iron, lead, molybdenum, nickel, thallium, tungsten, uranium, vanadium and zinc. (author)

  18. Uranium chloride extraction of transuranium elements from LWR fuel

    International Nuclear Information System (INIS)

    Miller, W.E.; Ackerman, J.P.; Battles, J.E.; Johnson, T.R.; Pierce, R.D.

    1992-01-01

    A process of separating transuranium actinide values from uranium values present in spent nuclear oxide fuels containing rare earth and noble metal fission products as well as other fission products is disclosed. The oxide fuel is reduced with Ca metal in the presence of Ca chloride and a U-Fe alloy which is liquid at about 800 C to dissolve uranium metal and the noble metal fission product metals and transuranium actinide metals and rare earth fission product metals leaving Ca chloride having CaO and fission products of alkali metals and the alkali earth metals and iodine dissolved therein. The Ca chloride and CaO and the fission products contained therein are separated from the U-Fe alloy and the metal values dissolved therein. The U-Fe alloy having dissolved therein reduced metals from the spent nuclear fuel is contacted with a mixture of one or more alkali metal or alkaline earth metal halides selected from the class consisting of alkali metal or alkaline earth metal and Fe or U halide or a combination thereof to transfer transuranium actinide metals and rare earth metals to the halide salt leaving the uranium and some noble metal fission products in the U-Fe alloy and thereafter separating the halide salt and the transuranium metals dissolved therein from the U-Fe alloy and the metals dissolved therein. 1 figure

  19. Uranium mining in Australia

    International Nuclear Information System (INIS)

    Mackay, G.A.

    1978-01-01

    Western world requirements for uranium based on increasing energy consumption and a changing energy mix, will warrant the development of Australia's resources. By 1985 Australian mines could be producing 9500 tonnes of uranium oxide yearly and by 1995 the export value from uranium could reach that from wool. In terms of benefit to the community the economic rewards are considerable but, in terms of providing energy to the world, Australias uranium is vital

  20. Bicarbonate leaching of uranium

    International Nuclear Information System (INIS)

    Mason, C.

    1998-01-01

    The alkaline leach process for extracting uranium from uranium ores is reviewed. This process is dependent on the chemistry of uranium and so is independent on the type of mining system (conventional, heap or in-situ) used. Particular reference is made to the geochemical conditions at Crownpoint. Some supporting data from studies using alkaline leach for remediation of uranium-contaminated sites is presented

  1. Bicarbonate leaching of uranium

    Energy Technology Data Exchange (ETDEWEB)

    Mason, C.

    1998-12-31

    The alkaline leach process for extracting uranium from uranium ores is reviewed. This process is dependent on the chemistry of uranium and so is independent on the type of mining system (conventional, heap or in-situ) used. Particular reference is made to the geochemical conditions at Crownpoint. Some supporting data from studies using alkaline leach for remediation of uranium-contaminated sites is presented.

  2. Uranium in fossil bones

    International Nuclear Information System (INIS)

    Koul, S.L.

    1978-01-01

    An attempt has been made to determine the uranium content and thus the age of certain fossil bones Haritalyangarh (Himachal Pradesh), India. The results indicate that bones rich in apatite are also rich in uranium, and that the radioactivity is due to radionuclides in the uranium series. The larger animals apparently have a higher concentration of uranium than the small. The dating of a fossil jaw (elephant) places it in the Pleistocene. (Auth.)

  3. Method for converting uranium oxides to uranium metal

    International Nuclear Information System (INIS)

    Duerksen, W.K.

    1988-01-01

    A method for converting uranium oxide to uranium metal is described comprising the steps of heating uranium oxide in the presence of a reducing agent to a temperature sufficient to reduce the uranium oxide to uranium metal and form a heterogeneous mixture of a uranium metal product and oxide by-products, heating the mixture in a hydrogen atmosphere at a temperature sufficient to convert uranium metal in the mixture to uranium hydride, cooling the resulting uranium hydride-containing mixture to a temperature sufficient to produce a ferromagnetic transition in the uranium hydride, magnetically separating the cooled uranium hydride from the mixture, and thereafter heating the separated uranium hydride in an inert atmosphere to a temperature sufficient to convert the uranium hydride to uranium metal

  4. Experimental measurement of fission fragments paths in uranium gold, molybdenum, zirconium and silicon; Mesure experimentale des parcours des fragments de fission dans l'uranium, l'or, le molybdene, le zirconium et le silicium

    Energy Technology Data Exchange (ETDEWEB)

    Faraggi, H; Garin-Bonnet, A [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1958-07-01

    The measurement of total number of fissiongments emerging from an homogeneous, thick alloy composed of uranium plus another element (the concentration of uranium being known) allows to obtain the range of the fragments in this alloy. By varying the concentration, the range of the fragments in uranium and in the other element can be deduced. (author)Fren. [French] La mesure du nombre total de fragments de fission sortant d'un alliage homogene epais d'uranium et d'un autre element, pour lequel la concentration en uranium est donnee, permet la mesure du parcours des fragments dans cet alliage. En faisant varier la concentration, on peut deduire de ces mesures le parcours des fragments dans l'uranium et dans l'autre element. (auteur)

  5. Microbial accumulation of uranium

    International Nuclear Information System (INIS)

    Zhang Wei; Dong Faqin; Dai Qunwei

    2005-01-01

    The mechanism of microbial accumulation of uranium and the effects of some factors (including pH, initial uranium concentration, pretreatment of bacteria, and so on) on microbial accumulation of uranium are discussed briefly. The research direction and application prospect are presented. (authors)

  6. Uranium energy dependence

    International Nuclear Information System (INIS)

    Erkes, P.

    1981-06-01

    Uranium supply and demand as projected by the Uranium Institute is discussed. It is concluded that for the industrialized countries, maximum energy independence is a necessity. Hence it is necessary to achieve assurance of supply for uranium used in thermal power reactors in current programs and eventually to move towards breeders

  7. Australian uranium today

    International Nuclear Information System (INIS)

    Fisk, B.

    1978-01-01

    The subject is covered in sections, entitled: Australia's resources; Northern Territory uranium in perspective; the government's decision [on August 25, 1977, that there should be further development of uranium under strictly controlled conditions]; Government legislation; outlook [for the Australian uranium mining industry]. (U.K.)

  8. Uranium resources, 1983

    International Nuclear Information System (INIS)

    1983-01-01

    The specific character of uranium as energy resources, the history of development of uranium resources, the production and reserve of uranium in the world, the prospect regarding the demand and supply of uranium, Japanese activity of exploring uranium resources in foreign countries and the state of development of uranium resources in various countries are reported. The formation of uranium deposits, the classification of uranium deposits and the reserve quantity of each type are described. As the geological environment of uranium deposits, there are six types, that is, quartz medium gravel conglomerate deposit, the deposit related to the unconformity in Proterozoic era, the dissemination type magma deposit, pegmatite deposit and contact deposit in igneaus rocks and metamorphic rocks, vein deposit, sandstone type deposit and the other types of deposit. The main features of respective types are explained. The most important uranium resources in Japan are those in the Tertiary formations, and most of the found reserve belongs to this type. The geological features, the state of yield and the scale of the deposits in Ningyotoge, Tono and Kanmon Mesozoic formation are reported. Uranium minerals, the promising districts in the world, and the matters related to the exploration and mining of uranium are described. (Kako, I.)

  9. Recycling of reprocessed uranium

    International Nuclear Information System (INIS)

    Randl, R.P.

    1987-01-01

    Since nuclear power was first exploited in the Federal Republic of Germany, the philosophy underlying the strategy of the nuclear fuel cycle has been to make optimum use of the resource potential of recovered uranium and plutonium within a closed fuel cycle. Apart from the weighty argument of reprocessing being an important step in the treatment and disposal of radioactive wastes, permitting their optimum ecological conditioning after the reprocessing step and subsequent storage underground, another argument that, no doubt, carried weight was the possibility of reducing the demand of power plants for natural uranium. In recent years, strategies of recycling have emerged for reprocessed uranium. If that energy potential, too, is to be exploited by thermal recycling, it is appropriate to choose a slightly different method of recycling from the one for plutonium. While the first generation of reprocessed uranium fuel recycled in the reactor cuts down natural uranium requirement by some 15%, the recycling of a second generation of reprocessed, once more enriched uranium fuel helps only to save a further three per cent of natural uranium. Uranium of the second generation already carries uranium-232 isotope, causing production disturbances, and uranium-236 isotope, causing disturbances of the neutron balance in the reactor, in such amounts as to make further fabrication of uranium fuel elements inexpedient, even after mixing with natural uranium feed. (orig./UA) [de

  10. Uranium absorption study pile; Empilement pour le controle de l'absorption de l'uranium

    Energy Technology Data Exchange (ETDEWEB)

    Raievski, V; Sautiez, B [Commissariat a l' Energie Atomique, Saclay (France).Centre d' Etudes Nucleaires

    1959-07-01

    The report describes a pile designed to measure the absorption of fuel slugs. The pile is of graphite and comprises a central section composed of uranium rods in a regular lattice. RaBe sources and BF{sub 3} counters are situated on either side of the center. A given uranium charge is compared with a specimen charge of about 560 kg, and the difference in absorption between the two noted. The sensitivity of the equipment will detect absorption variations of about a few ppm boron (10{sup -6} boron per gr. of uranium) or better. (author) [French] Nous decrivons un dispositif permettant de mesurer l'absorption des elements combustibles d'une pile. Ce dispositif est constitue par un empilement de graphite dont la region centrale est formee par un reseau regulier de barres d'uranium. Des sources de RaBe et des compteurs a BF{sub 3} sont places de part et d'autre de cette region. En comparant un chargement d'uranium a un chargement etalon d'environ 560 kg, on peut determiner la difference d'absorption entre ces deux chargements. La sensibilite permettrait de deceler une variation d'absorption de l'ordre du ppm de bore (10{sup -6} g de bore par gramme d'uranium) et peut-etre mieux. (auteur)

  11. VANADIUM ALLOYS

    Science.gov (United States)

    Smith, K.F.; Van Thyne, R.J.

    1959-05-12

    This patent deals with vanadium based ternary alloys useful as fuel element jackets. According to the invention the ternary vanadium alloys, prepared in an arc furnace, contain from 2.5 to 15% by weight titanium and from 0.5 to 10% by weight niobium. Characteristics of these alloys are good thermal conductivity, low neutron capture cross section, good corrosion resistance, good welding and fabricating properties, low expansion coefficient, and high strength.

  12. High loading uranium plate

    International Nuclear Information System (INIS)

    Wiencek, T.C.; Domagala, R.F.; Thresh, H.R.

    1990-01-01

    Two embodiments of a high uranium fuel plate are disclosed which contain a meat comprising structured uranium compound confined between a pari of diffusion bonded ductile metal cladding plates uniformly covering the meat, the meat hiving a uniform high fuel loading comprising a content of uranium compound greater than about 45 Vol. % at a porosity not greater than about 10 Vol. %. In a first embodiment, the meat is a plurality of parallel wires of uranium compound. In a second embodiment, the meat is a dispersion compact containing uranium compound. The fuel plates are fabricated by a hot isostatic pressing process

  13. PROCESS OF RECOVERING URANIUM

    Science.gov (United States)

    Carter, J.M.; Larson, C.E.

    1958-10-01

    A process is presented for recovering uranium values from calutron deposits. The process consists in treating such deposits to produce an oxidlzed acidic solution containing uranium together with the following imparities: Cu, Fe, Cr, Ni, Mn, Zn. The uranium is recovered from such an impurity-bearing solution by adjusting the pH of the solution to the range 1.5 to 3.0 and then treating the solution with hydrogen peroxide. This results in the precipitation of uranium peroxide which is substantially free of the metal impurities in the solution. The peroxide precipitate is then separated from the solution, washed, and calcined to produce uranium trioxide.

  14. Phase transformation of metastable cubic γ-phase in U-Mo alloys

    International Nuclear Information System (INIS)

    Sinha, V.P.; Hegde, P.V.; Prasad, G.J.; Dey, G.K.; Kamath, H.S.

    2010-01-01

    Over the past decade considerable efforts have been put by many fuel designers to develop low enriched uranium (LEU 235 ) base U-Mo alloy as a potential fuel for core conversion of existing research and test reactors which are running on high enriched uranium (HEU > 85%U 235 ) fuel and also for the upcoming new reactors. U-Mo alloy with minimum 8 wt% molybdenum shows excellent metastability with cubic γ-phase in cast condition. However, it is important to characterize the decomposition behaviour of metastable cubic γ-uranium in its equilibrium products for in reactor fuel performance point of view. The present paper describes the phase transformation behaviour of cubic γ-uranium phase in U-Mo alloys with three different molybdenum compositions (i.e. 8 wt%, 9 wt% and 10 wt%). U-Mo alloys were prepared in an induction melting furnace and characterized by X-ray diffraction (XRD) method for phase determination. Microstructures were developed for samples in as cast condition. The alloys were hot rolled in cubic γ-phase to break the cast structure and then they were aged at 500 o C for 68 h and 240 h, so that metastable cubic γ-uranium will undergo eutectoid decomposition to form equilibrium phases of orthorhombic α-uranium and body centered tetragonal U 2 Mo intermetallic compound. U-Mo alloy samples with different ageing history were then characterized by XRD for phase and development of microstructure.

  15. Method for converting uranium oxides to uranium metal

    Science.gov (United States)

    Duerksen, Walter K.

    1988-01-01

    A process is described for converting scrap and waste uranium oxide to uranium metal. The uranium oxide is sequentially reduced with a suitable reducing agent to a mixture of uranium metal and oxide products. The uranium metal is then converted to uranium hydride and the uranium hydride-containing mixture is then cooled to a temperature less than -100.degree. C. in an inert liquid which renders the uranium hydride ferromagnetic. The uranium hydride is then magnetically separated from the cooled mixture. The separated uranium hydride is readily converted to uranium metal by heating in an inert atmosphere. This process is environmentally acceptable and eliminates the use of hydrogen fluoride as well as the explosive conditions encountered in the previously employed bomb-reduction processes utilized for converting uranium oxides to uranium metal.

  16. Uranium speciation in plants

    International Nuclear Information System (INIS)

    Guenther, A.; Bernhard, G.; Geipel, G.; Reich, T.; Rossberg, A.; Nitsche, H.

    2003-01-01

    Detailed knowledge of the nature of uranium complexes formed after the uptake by plants is an essential prerequisite to describe the migration behavior of uranium in the environment. This study focuses on the determination of uranium speciation after uptake of uranium by lupine plants. For the first time, time-resolved laser-induced fluorescence spectroscopy and X-ray absorption spectroscopy were used to determine the chemical speciation of uranium in plants. Differences were detected between the uranium speciation in the initial solution (hydroponic solution and pore water of soil) and inside the lupine plants. The oxidation state of uranium did not change and remained hexavalent after it was taken up by the lupine plants. The chemical speciation of uranium was identical in the roots, shoot axis, and leaves and was independent of the uranium speciation in the uptake solution. The results indicate that the uranium is predominantly bound as uranyl(VI) phosphate to the phosphoryl groups. Dandelions and lamb's lettuce showed uranium speciation identical to lupine plants. (orig.)

  17. Refining U-Zr-Nb alloys by remelting

    International Nuclear Information System (INIS)

    Aguiar, B.M.; Kniess, C.T.; Riella, H.G.; Ferraz, W.B.

    2011-01-01

    The high density U-Zr-Nb and U-Nb uranium-based alloys can be employed as nuclear fuel in a PWR reactor due to their high density and nuclear properties. These alloys can stabilize the gamma phase, however, according to TTT diagrams, at the working temperature of a PWR reactor, all gamma phase transforms to α'' phase in a few hours. To avoid this kind of transformation during the nuclear reactor operation, the U-Zr-Nb alloy and U-Nn are used in α'' phase. The stability of α'' phase depends on the alloy composition and cooling rate. The alloy homogenization has to be very effective to eliminate precipitates rich in Zr and Nb to avoid changes in the alloying elements contents in the matrix. The homogenization was obtained by remelting the alloy and keeping it in the liquid state for enough time to promote floating of the precipitates (usually carbides, less dense) and leaving the matrix free of precipitates. However, this floating by density difference may result in segregation between the alloying elements (Nb and Zr, at the top) and uranium (at the bottom). The homogenized alloys were characterized in terms of metallographic techniques, optical microscopy, scanning electronic microscopy, EDS and X-ray diffraction. In this paper, it is shown that the contents of Zr and Nb at the bottom and at the top of the matrix are constant. (author)

  18. Refining U-Zr-Nb alloys by remelting

    Energy Technology Data Exchange (ETDEWEB)

    Aguiar, B.M.; Kniess, C.T.; Riella, H.G., E-mail: bmaguiar@ipen.b [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil); Ferraz, W.B. [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

    2011-07-01

    The high density U-Zr-Nb and U-Nb uranium-based alloys can be employed as nuclear fuel in a PWR reactor due to their high density and nuclear properties. These alloys can stabilize the gamma phase, however, according to TTT diagrams, at the working temperature of a PWR reactor, all gamma phase transforms to {alpha}'' phase in a few hours. To avoid this kind of transformation during the nuclear reactor operation, the U-Zr-Nb alloy and U-Nn are used in {alpha}'' phase. The stability of {alpha}'' phase depends on the alloy composition and cooling rate. The alloy homogenization has to be very effective to eliminate precipitates rich in Zr and Nb to avoid changes in the alloying elements contents in the matrix. The homogenization was obtained by remelting the alloy and keeping it in the liquid state for enough time to promote floating of the precipitates (usually carbides, less dense) and leaving the matrix free of precipitates. However, this floating by density difference may result in segregation between the alloying elements (Nb and Zr, at the top) and uranium (at the bottom). The homogenized alloys were characterized in terms of metallographic techniques, optical microscopy, scanning electronic microscopy, EDS and X-ray diffraction. In this paper, it is shown that the contents of Zr and Nb at the bottom and at the top of the matrix are constant. (author)

  19. 31 CFR 540.317 - Uranium feed; natural uranium feed.

    Science.gov (United States)

    2010-07-01

    ... 31 Money and Finance: Treasury 3 2010-07-01 2010-07-01 false Uranium feed; natural uranium feed... (Continued) OFFICE OF FOREIGN ASSETS CONTROL, DEPARTMENT OF THE TREASURY HIGHLY ENRICHED URANIUM (HEU) AGREEMENT ASSETS CONTROL REGULATIONS General Definitions § 540.317 Uranium feed; natural uranium feed. The...

  20. The Effect of Cold Rolling on the Hydrogen Susceptibility of 5083 Aluminum Alloy

    Directory of Open Access Journals (Sweden)

    E.P. Georgiou

    2017-10-01

    Full Text Available This work focuses in investigating the effect of cold deformation on the cathodic hydrogen charging of 5083 aluminum alloy. The aluminium alloy was submitted to a cold rolling process, until the average thickness of the specimens was reduced by 7% and 15%, respectively. A study of the structure, microhardness, and tensile properties of the hydrogen charged aluminium specimens, with and without cold rolling, indicated that the cold deformation process led to an increase of hydrogen susceptibility of this aluminum alloy.

  1. Nonswelling alloy

    Science.gov (United States)

    Harkness, S.D.

    1975-12-23

    An aluminum alloy containing one weight percent copper has been found to be resistant to void formation and thus is useful in all nuclear applications which currently use aluminum or other aluminum alloys in reactor positions which are subjected to high neutron doses.

  2. Nonswelling alloy

    International Nuclear Information System (INIS)

    Harkness, S.D.

    1975-01-01

    An aluminum alloy containing one weight percent copper has been found to be resistant to void formation and thus is useful in all nuclear applications which currently use aluminum or other aluminum alloys in reactor positions which are subjected to high neutron doses

  3. A slant type shape memory alloy

    International Nuclear Information System (INIS)

    Kanada, T.; Enokizono, M.

    2000-01-01

    A heat-treated Fe-based shape memory alloy (SMA) has compatible properties, magnetization and shape memory effect (SME). Since SME depends on the heat treatment conditions (temperature and time), we produced a slant-type SMA that has a gradient SME value in the longitudinal direction of the specimen. It is obvious that sheet specimen is superior to wire because the value of SME as a slant SME shows greater efficiency than that of wire

  4. Mechanical Properties of Spray Cast 7XXX Series Aluminium Alloys

    OpenAIRE

    SALAMCI, Elmas

    2014-01-01

    Mechanical properties of spray deposited and extruded 7xxx series aluminium alloys were investigated in peak aged condition. To study the influence of Zn additions on the mechanical behaviour of spray deposited materials, three alloy compositions were selected, namely: SS70 (11.5% Zn), N707 (10.9% Zn) and 7075 (5.6% Zn). After ageing treatment, notched and unnotched specimens of spray deposited alloys were subjected to tensile tests at room temperature. Experimental results showed...

  5. Investigation of americium-241 metal alloys for target applications

    International Nuclear Information System (INIS)

    Conner, W.V.; Rockwell International Corp., Golden, CO

    1982-01-01

    Several 241 Am metal alloys have been investigated for possible use in the Lawrence Livermore National Laboratory Radiochemical Diagnostic Tracer Program. Several properties were desired for an alloy to be useful for tracer program applications. A suitable alloy would have a fairly high density, be ductile, homogeneous and easy to prepare. Alloys investigated have included uranium-americium, aluminium-americium, and cerium-americium. Uranium-americium alloys with the desired properties proved to be difficult to prepare, and work with this alloy was discontinued. Aluminium-americium alloys were much easier to prepare, but the alloy consisted of an aluminium-americium intermetallic compound (AmAl 4 ) in an aluminum matrix. This alloy could be cast and formed into shapes, but the low density of aluminum, and other problems, made the alloy unsuitable for the intended application. Americium metal was found to have a high solid solubility in cerium and alloys prepared from these two elements exhibited all of the properties desired for the tracer program application. Cerium-americium alloys containing up to 34 wt% americium have been prepared using both co-melting and co-reduction techniques. The latter technique involves co-reduction of cerium tetrafluoride and americium tetrafluoride with calcium metal in a sealed reduction vessel. Casting techniques have been developed for preparing up to eight 2.2 cm (0.87 in) diameter disks in a single casting, and cerium-americium metal alloy disks containing from 10 to 25 wt% 241 Am have been prepared using these techniques. (orig.)

  6. 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

  7. 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

  8. Procedure for Uranium-Molybdenum Density Measurements and Porosity Determination

    Energy Technology Data Exchange (ETDEWEB)

    Prabhakaran, Ramprashad [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Devaraj, Arun [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Joshi, Vineet V. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Lavender, Curt A. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2016-08-13

    The purpose of this document is to provide guidelines for preparing uranium-molybdenum (U-Mo) specimens, performing density measurements, and computing sample porosity. Typical specimens (solids) will be sheared to small rectangular foils, disks, or pieces of metal. A mass balance, solid density determination kit, and a liquid of known density will be used to determine the density of U-Mo specimens using the Archimedes principle. A standard test weight of known density would be used to verify proper operation of the system. By measuring the density of a U-Mo sample, it is possible to determine its porosity.

  9. 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

  10. Nature of negative microplastic deformation in alloys

    International Nuclear Information System (INIS)

    Palatnik, L.S.; Ivantsov, V.I.; Kagan, Ya.I.; Papirov, I.I.; Fat'yanova, N.B.; AN Ukrainskoj SSR, Kharkov. Fiziko-Tekhnicheskij Inst.)

    1985-01-01

    The paper deals with investigation of microplastic deformation of corrosion resistant aging 40KhNYU alloy and the study of physical nature of negative microdeformation in this alloy under tension. Investigation of microplasticity of 40KhNYU alloy was conducted by the method of mechanostatic hysteresis using resistance strain gauge for measuring stresses and deformations. Microplasticity curves for 40KhNYU alloy were obtained. They represent the result of competition between usual (positive) microdeformation and phase (negative) deformation under tensile effect on the alloy. It was established that the negative microdeformation increment occurs during secondary aging of the phase precipitated from initial supersat urated solid solution (primary decomposition product). This phase decomposes under tension with disperse phase precipitation which promotes decreasing its specific volume and specimen volume as a whole

  11. Effect of neutron irradiation on vanadium alloys

    International Nuclear Information System (INIS)

    Braski, D.N.

    1986-01-01

    Neutron-irradiated vanadium alloys were evaluated for their susceptibility to irradiation hardening, helium embrittlement, swelling, and residual radioactivity, and the results were compared with those for the austenitic and ferritic stainless steels. The VANSTAR-7 and V-15Cr-5Ti alloys showed the greatest hardening between 400 and 600 0 C while V-3Ti-1Si and V-20Ti had lower values that were comparable to those of ferritic steels. The V-15Cr-5Ti and VANSTAR-7 alloys were susceptible to helium embrittlement caused by the combination of weakened grain boundaries and irradiation-hardened grain matrices. Specimen fractures were entirely intergranular in the most severe instances of embrittlement. The V-3Ti-1Si and V-20Ti alloys were more resistant to helium embrittlement. Except for VANSTAR-7 irradiated to 40 dpa at 520 0 C, all of the vanadium alloys exhibited low swelling that was similar to the ferritic steels. Swelling was greater in specimens that were preimplanted with helium using the tritium trick. The vanadium alloys clearly exhibit lower residual radioactivity after irradiation than the ferrous alloys

  12. Effect of neutron irradiation on vanadium alloys

    Energy Technology Data Exchange (ETDEWEB)

    Braski, D.N.

    1986-01-01

    Neutron-irradiated vanadium alloys were evaluated for their susceptibility to irradiation hardening, helium embrittlement, swelling, and residual radioactivity, and the results were compared with those for the austenitic and ferritic stainless steels. The VANSTAR-7 and V-15Cr-5Ti alloys showed the greatest hardening between 400 and 600/sup 0/C while V-3Ti-1Si and V-20Ti had lower values that were comparable to those of ferritic steels. The V-15Cr-5Ti and VANSTAR-7 alloys were susceptible to helium embrittlement caused by the combination of weakened grain boundaries and irradiation-hardened grain matrices. Specimen fractures were entirely intergranular in the most severe instances of embrittlement. The V-3Ti-1Si and V-20Ti alloys were more resistant to helium embrittlement. Except for VANSTAR-7 irradiated to 40 dpa at 520/sup 0/C, all of the vanadium alloys exhibited low swelling that was similar to the ferritic steels. Swelling was greater in specimens that were preimplanted with helium using the tritium trick. The vanadium alloys clearly exhibit lower residual radioactivity after irradiation than the ferrous alloys.

  13. Collagen Quantification in Tissue Specimens.

    Science.gov (United States)

    Coentro, João Quintas; Capella-Monsonís, Héctor; Graceffa, Valeria; Wu, Zhuning; Mullen, Anne Maria; Raghunath, Michael; Zeugolis, Dimitrios I

    2017-01-01

    Collagen is the major extracellular protein in mammals. Accurate quantification of collagen is essential in the biomaterials (e.g., reproducible collagen scaffold fabrication), drug discovery (e.g., assessment of collagen in pathophysiologies, such as fibrosis), and tissue engineering (e.g., quantification of cell-synthesized collagen) fields. Although measuring hydroxyproline content is the most widely used method to quantify collagen in biological specimens, the process is very laborious. To this end, the Sircol™ Collagen Assay is widely used due to its inherent simplicity and convenience. However, this method leads to overestimation of collagen content due to the interaction of Sirius red with basic amino acids of non-collagenous proteins. Herein, we describe the addition of an ultrafiltration purification step in the process to accurately determine collagen content in tissues.

  14. Study of elementary mechanisms of creep in uranium as a function of temperature (150 deg. to 760 deg. C) by activation energy measurements

    International Nuclear Information System (INIS)

    Grenier, P.

    1966-06-01

    Creep tests were carried out on single crystals and polycrystalline specimens of uranium in both the α and β phases over the temperature range 150 - 760 deg. C. The determination of the activation energy for creep and the study of its variation with temperature made it possible to distinguish various temperature ranges in which one or more elementary mechanisms govern deformation. Micrographic observations after creep and the study of the variation of creep-rate with load support the conclusions. The creep behavior of single crystals is identical with that of polycrystalline material below 325 deg. C. From 325 deg. C to one upper limiting temperature whose value depends on the purity and previous history of the metal, the creep deformation of uranium is controlled by cross-slip. From this limiting temperature up to 520 deg. C, the creep of uranium involves two independent mechanisms operating simultaneously, the movement of screw dislocation by cross-slip and the climbing of edge dislocations out of their slip plane. Between 520 deg. C and the α - β transformation temperature creep in polycrystals is governed by the climb of edge dislocations out of their slip planes, by a pile up mechanism in the case of primary creep and by dipole annihilation in the case of secondary creep. In single crystals creep is dependent on the climb of edge dislocations into pre-existent sub-boundaries and their subsequent rearrangement within these boundaries. In the β phase the creep of polycrystals is governed by the diffusional climb of edge dislocations. Between 450 and 630 deg. C small alloy additions of molybdenum modify the creep characteristics of uranium although the deformation mechanisms involved are analogous to those in the pure metal. (author) [fr

  15. Uranium of Kazakhstan

    International Nuclear Information System (INIS)

    Tsalyuk, Yu.; Gurevich, D.

    2000-01-01

    Over 25 % of the world's uranium reserves are concentrated in Kazakhstan. So, the world's largest Shu-Sarysu uranium province is situated on southern Kazakhstan, with resources exceeding 1 billion tonnes of uranium. No less, than 3 unique deposits with resources exceeding 100,000 tonnes are situated here. From the economic point of view the most important thing is that these deposits are suitable for in-situ leaching, which is the cheapest, environmentally friendly and most efficient method available for uranium extracting. In 1997 the Kazatomprom National Joint-Stock Company united all Kazakhstan's uranium enterprises (3 mine and concentrating plants, Volkovgeologiya Joint-Stock Company and the Ulbinskij Metallurgical plant). In 1998 uranium production came to 1,500 tonnes (860 kg in 1997). In 1999 investment to the industry were about $ 30 million. Plans for development of Kazakhstan's uranium industry provide a significant role for foreign partners. At present, 2 large companies (Comeco (Canada), Cogema (France) working in Kazakhstan. Kazakatomprom continues to attract foreign investors. The company's administration announced that in that in next year they have plan to make a radical step: to sell 67 % of stocks to strategic investors (at present 100 % of stocks belongs to state). Authors of the article regard, that the Kazakhstan's uranium industry still has significant reserves to develop. Even if the scenario for the uranium industry could be unfavorable, uranium production in Kazakhstan may triple within the next three to four years. The processing of uranium by the Ulbinskij Metallurgical Plant and the production of some by-products, such as rhenium, vanadium and rare-earth elements, may provide more profits. Obviously, the sale of uranium (as well as of any other reserves) cannot make Kazakhstan a prosperous country. However, country's uranium industry has a god chance to become one of the most important and advanced sectors of national economy

  16. Notch effects in uniaxial tension specimens

    International Nuclear Information System (INIS)

    Delph, T.J.

    1979-03-01

    Results of a literature survey on the effect of notches on the time-dependent failure of uniaxial tension specimens at elevated temperatures are presented. Particular attention is paid to the failure of notched specimens containing weldments

  17. Measurements and Counts for Notacanthidae Specimens

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — Taxonomic data were collected for specimens of deep-sea spiny eels (Notacanthidae) from the Hawaiian Ridge by Bruce C. Mundy. Specimens were collected off the north...

  18. Hydrogen embrittlement of titanium and its alloys - a literature review

    International Nuclear Information System (INIS)

    Aho-Mantila, I.; Haemaelaeinen, H.

    1986-05-01

    Hydrogen embrittlement data of titanium and its alloys is reviewed. Especially the results obtained in spent nuclear fuel repository conditions with commercially pure titanium and TiCode-12 alloy are examined. The results show that the mechanical properties of titanium are not much affected by hydrogen when tested by smooth specimens. Much greater effects can be expected with notched fracture mechanics specimens. However, only limeted data is available. Hydrogen distribution in titanium is affected by stress, alloy composition and temperature gradients. In order to model the hydrogen-induced crack growth in titanium much more mechanistic work is needed especially to understand the behaviour of hydrogen in crack tip stress field. (author)

  19. Theoretical Model for Volume Fraction of UC, 235U Enrichment, and Effective Density of Final U 10Mo Alloy

    Energy Technology Data Exchange (ETDEWEB)

    Devaraj, Arun [Pacific Northwest National Lab. (PNNL), Richland, WA (United States). Environmental Molecular Sciences Lab. (EMSL); Prabhakaran, Ramprashad [Pacific Northwest National Lab. (PNNL), Richland, WA (United States). Environmental Molecular Sciences Lab. (EMSL); Joshi, Vineet V. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States). Environmental Molecular Sciences Lab. (EMSL); Hu, Shenyang Y. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States). Environmental Molecular Sciences Lab. (EMSL); McGarrah, Eric J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States). Environmental Molecular Sciences Lab. (EMSL); Lavender, Curt A. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States). Environmental Molecular Sciences Lab. (EMSL)

    2016-04-12

    The purpose of this document is to provide a theoretical framework for (1) estimating uranium carbide (UC) volume fraction in a final alloy of uranium with 10 weight percent molybdenum (U-10Mo) as a function of final alloy carbon concentration, and (2) estimating effective 235U enrichment in the U-10Mo matrix after accounting for loss of 235U in forming UC. This report will also serve as a theoretical baseline for effective density of as-cast low-enriched U-10Mo alloy. Therefore, this report will serve as the baseline for quality control of final alloy carbon content

  20. Comparison measure of natural radioactivity in environment specimen using HPGe and NaI(Tl) γ-ray spectrometer

    International Nuclear Information System (INIS)

    Zhou Chunlin; Han Feng; Li Tiantuo; Ma Wenyan; Di Yuming; Guo Huiping; Wu Yuelei

    2000-01-01

    The author reports the comparison results on natural radioactive nuclide contents of soil specimen from an uranium diggings with HPGe and NaI(Tl) γ-ray spectrometer. Relative method and athwart matrix method are used to analyze natural radioactive nuclide contents in samples of soil. The results are compared and are proven to be in accordance with each other