Microstructure Modeling of 3rd Generation Disk Alloy

Science.gov (United States)

Jou, Herng-Jeng

2008-01-01

The objective of this initiative, funded by NASA's Aviation Safety Program, is to model, validate, and predict, with high fidelity, the microstructural evolution of third-generation high-refractory Ni-based disc superalloys during heat treating and service conditions. This initiative is a natural extension of the DARPA-AIM (Accelerated Insertion of Materials) initiative with GE/Pratt-Whitney and with other process simulation tools. Strong collaboration with the NASA Glenn Research Center (GRC) is a key component of this initiative and the focus of this program is on industrially relevant disk alloys and heat treatment processes identified by GRC. Employing QuesTek s Computational Materials Dynamics technology and PrecipiCalc precipitation simulator, physics-based models are being used to achieve high predictive accuracy and precision. Combining these models with experimental data and probabilistic analysis, "virtual alloy design" can be performed. The predicted microstructures can be optimized to promote desirable features and concurrently eliminate nondesirable phases that can limit the reliability and durability of the alloys. The well-calibrated and well-integrated software tools that are being applied under the proposed program will help gas turbine disk alloy manufacturers, processing facilities, and NASA, to efficiently and effectively improve the performance of current and future disk materials.

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

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

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

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.)

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

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.

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

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)

Development of materials and process technology for dual alloy disks

Science.gov (United States)

Marder, J. M.; Kortovich, C. S.

1981-01-01

Techniques for the preparation of dual alloy disks were developed and evaluated. Four material combinations were evaluated in the form of HIP consolidated and heat treated cylindrical and plate shapes in terms of elevated temperature tensile, stress rupture and low cycle fatigue properties. The process evaluation indicated that the pe-HIP AF-115 rim/loose powder Rene 95 hub combination offered the best overall range of mechanical properties for dual disk applications. The feasibility of this dual alloy concept for the production of more complex components was demonstrated by the scale up fabrication of a prototype CFM-56 disk made from this AF-115/Rene 95 combination. The hub alloy ultimate tensile strength was approximately 92 percent of the program goal of 1520 MPa (220 ksi) at 480 C (900 F) and the rim alloy stress rupture goal of 300 hours at 675 C (1250 F)/925 MPa (134 ksi) was exceeded by 200 hours. The low cycle fatigue properties were equivalent to those exhibited by HIP and heat treated alloys. There was an absence of rupture notch sensitivity in both alloys. The joint tensile properties were approximately 85 percent of the weaker of the two materials (Rene 95) and the stress rupture properties were equivalent to those of the weaker of the two materials (Rene 95).

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

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

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

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.)

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)

Yield surface investigation of alloys during model disk spin tests

Directory of Open Access Journals (Sweden)

E. P. Kuzmin

2014-01-01

Full Text Available Gas-turbine engines operate under heavy subsequently static loading conditions. Disks of gas-turbine engine are high loaded parts of irregular shape having intensive stress concentrators wherein a 3D stress strain state occurs. The loss of load-carrying capability or burst of disk can lead to severe accident or disaster. Therefore, development of methods to assess deformations and to predict burst is one of the most important problems.Strength assessment approaches are used at all levels of engine creation. In recent years due to actively developing numerical method, particularly FEA, it became possible to investigate load-carrying capability of irregular shape disks, to use 3D computing schemes including flow theory and different options of force and deformation failure criteria. In spite of a wide progress and practical use of strength assessment approaches, there is a lack of detailed research data on yield surface of disk alloys. The main purpose of this work is to validate the use of basis hypothesis of flow theory and investigate the yield surface of disk alloys during the disks spin test.The results of quasi-static numerical simulation of spin tests of model disk made from high-temperature forged alloy are presented. To determine stress-strain state of disk during loading finite element analysis is used. Simulation of elastic-plastic strain fields was carried out using incremental theory of plasticity with isotropic hardening. Hardening function was taken from the results of specimens tensile test. Specimens were cut from a sinkhead of model disk. The paper investigates the model sensitivity affected by V.Mises and Tresca yield criteria as well as the Hosford model. To identify the material model parameters the eddy current sensors were used in the experimental approach to measure rim radial displacements during the load-unload of spin test. The results of calculation made using different material models were compared with the

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)

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)

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)

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

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)

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)

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

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

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)

Examination of disks from the IPNS depleted uranium target

International Nuclear Information System (INIS)

Strain, R.V.; Carpenter, J.M.

1995-10-01

This report describes the results of examining the Zircaloy-2 clad depleted uranium disks from the Intense Pulse Neutron Source (IPNS) Target. That target operated from August, 1981 to June, 1988 and from September, 1991 to September, 1992 at 450 MeV, pulsing at 30 Hz with a time average proton current of about 15 microA. The target was removed from service when the presence of fission products ( 135 Xe) in the coolant cover gas indicated a failure in the Zircaloy-2 cladding. Altogether, the target had absorbed about 240 mA hours of proton current, and endured between 50,000 and 100,000 thermal cycles. The purpose of the examination was to assess the condition of the disks and determine the cause of the cladding failure. The results of visual, gamma ray scanning, and destructive metallurgical examination of two disks are described

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)

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

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)

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

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)

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

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.)

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

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

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)

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

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.

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

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

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

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

Minaturized disk bend tests of neutron-irradiated path A type alloys

International Nuclear Information System (INIS)

Lee, M.; Sohn, D.S.; Grant, N.J.; Harling, O.K.

1983-01-01

Path A Prime Candidate Alloy (PCA) has been rapidly solidified and consoliated by extrusion. Twenty percent CW samples, precision TEM disks, 3 phi x 0.254 mm, were irradiated in the mixed flux of the Oak Ridge HFIR reactor up to approx. 8.5 dpa (360 appm He) and approx. 34 dpa (3100 appm He) at 300, 400, 500 and 600 0 C. Similar samples of conventionally processed PCA were also irradiated for comparison. Mechanical properties were characterized using a minaturized disk bend test (MDBT) developed at MIT. These tests indicate major decreases in strength and ductility especially for the 500 and 600 0 C irradiations. No major differences were found between this first version of a rapidly solidified and extruded PCA type alloy and conventionally processed PCA

Electrodeposition of milligram amounts of uranium on electropolished stainless steel disks

International Nuclear Information System (INIS)

Aggarwal, S.K.; Shah, P.M.; Duggal, R.K.; Jain, H.C.

1991-01-01

Investigations have been carried out for the electrodeposition of milligram amounts of uranium on electropolished stainless steel disks with an objective of preparing good quality sources for α-spectrometric studies on uranium. The parameters studied include the vatiation of electrodeposition yield as a function of voltage, time, distance between the cathode and anode, and the volume of 0.2M ammonium oxalate solution. The conditions selected for preparing good quality sources with nearly 100% yield were: electrodeposition voltage 25 V, time of deposition 15 min, volume of 0.2M ammonium oxalate solution in the cell 4 ml and a distance of 2 cm between the cathode and anode. The sources prepared using this method have been used successfully for the α-spectrometric determination of 234 U/ 238 U ratios in uranium samples. (author) 6 refs.; 4 figs

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

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

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

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

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

Development of high uranium-density fuels for use in research reactors

International Nuclear Information System (INIS)

Ugajin, Mitsuhiro; Akabori, Mitsuo; Itoh, Akinori

1996-01-01

The uranium silicide U 3 Si 2 possesses uranium density 11.3 gU/cm 3 with a congruent melting point of 1665degC, and is now successfully in use as a research reactor fuel. Another uranium silicide U 3 Si and U 6 Me-type uranium alloys (Me=Fe,Mn,Ni) have been chosen as new fuel materials because of the higher uranium densities 14.9 and 17.0 gU/cm 3 , respectively. Experiments were carried out to fabricate miniature aluminum-dispersion plate-type and aluminum-clad disk-type fuels by using the conventional picture-frame method and a hot-pressing technique, respectively. These included the above-mentioned new fuel materials as well as U 3 Si 2 . Totally 14 miniplates with uranium densities from 4.0 to 6.3 gU/cm 3 of fuel meat were prepared together with 28 disk-type fuel containing structurally-modified U 3 Si, and subjected to the neutron irradiation in JMTR (Japan Materials Testing Reactor). Some results of postirradiation examinations are presented. (author)

Powder metallurgy processing of high strength turbine disk alloys

Science.gov (United States)

Evans, D. J.

1976-01-01

Using vacuum-atomized AF2-1DA and Mar-M432 powders, full-scale gas turbine engine disks were fabricated by hot isostatically pressing (HIP) billets which were then isothermally forged using the Pratt & Whitney Aircraft GATORIZING forging process. While a sound forging was produced in the AF2-1DA, a container leak had occurred in the Mar-M432 billet during HIP. This resulted in billet cracking during forging. In-process control procedures were developed to identify such leaks. The AF2-1DA forging was heat treated and metallographic and mechanical property evaluation was performed. Mechanical properties exceeded those of Astroloy, one of the highest temperature capability turbine disk alloys presently used.

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.

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

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

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

Development of high uranium-density fuels for use in research reactors

Energy Technology Data Exchange (ETDEWEB)

Ugajin, Mitsuhiro; Akabori, Mitsuo; Itoh, Akinori [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

1996-02-01

The uranium silicide U{sub 3}Si{sub 2} possesses uranium density 11.3 gU/cm{sup 3} with a congruent melting point of 1665degC, and is now successfully in use as a research reactor fuel. Another uranium silicide U{sub 3}Si and U{sub 6}Me-type uranium alloys (Me=Fe,Mn,Ni) have been chosen as new fuel materials because of the higher uranium densities 14.9 and 17.0 gU/cm{sup 3}, respectively. Experiments were carried out to fabricate miniature aluminum-dispersion plate-type and aluminum-clad disk-type fuels by using the conventional picture-frame method and a hot-pressing technique, respectively. These included the above-mentioned new fuel materials as well as U{sub 3}Si{sub 2}. Totally 14 miniplates with uranium densities from 4.0 to 6.3 gU/cm{sup 3} of fuel meat were prepared together with 28 disk-type fuel containing structurally-modified U{sub 3}Si, and subjected to the neutron irradiation in JMTR (Japan Materials Testing Reactor). Some results of postirradiation examinations are presented. (author)

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)

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.)

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

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

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

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.

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)

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

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

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

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

Evaluation of powder metallurgy superalloy disk materials

Science.gov (United States)

Evans, D. J.

1975-01-01

A program was conducted to develop nickel-base superalloy disk material using prealloyed powder metallurgy techniques. The program included fabrication of test specimens and subscale turbine disks from four different prealloyed powders (NASA-TRW-VIA, AF2-1DA, Mar-M-432 and MERL 80). Based on evaluation of these specimens and disks, two alloys (AF2-1DA and Mar-M-432) were selected for scale-up evaluation. Using fabricating experience gained in the subscale turbine disk effort, test specimens and full scale turbine disks were formed from the selected alloys. These specimens and disks were then subjected to a rigorous test program to evaluate their physical properties and determine their suitability for use in advanced performance turbine engines. A major objective of the program was to develop processes which would yield alloy properties that would be repeatable in producing jet engine disks from the same powder metallurgy alloys. The feasibility of manufacturing full scale gas turbine engine disks by thermomechanical processing of pre-alloyed metal powders was demonstrated. AF2-1DA was shown to possess tensile and creep-rupture properties in excess of those of Astroloy, one of the highest temperature capability disk alloys now in production. It was determined that metallographic evaluation after post-HIP elevated temperature exposure should be used to verify the effectiveness of consolidation of hot isostatically pressed billets.

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)

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)

Forging of Advanced Disk Alloy LSHR

Science.gov (United States)

Gabb, Timothy P.; Gayda, John; Falsey, John

2005-01-01

The powder metallurgy disk alloy LSHR was designed with a relatively low gamma precipitate solvus temperature and high refractory element content to allow versatile heat treatment processing combined with high tensile, creep and fatigue properties. Grain size can be chiefly controlled through proper selection of solution heat treatment temperatures relative to the gamma precipitate solvus temperature. However, forging process conditions can also significantly influence solution heat treatment-grain size response. Therefore, it is necessary to understand the relationships between forging process conditions and the eventual grain size of solution heat treated material. A series of forging experiments were performed with subsequent subsolvus and supersolvus heat treatments, in search of suitable forging conditions for producing uniform fine grain and coarse grain microstructures. Subsolvus, supersolvus, and combined subsolvus plus supersolvus heat treatments were then applied. Forging and subsequent heat treatment conditions were identified allowing uniform fine and coarse grain microstructures.

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

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

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)

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.

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

Burst Testing and Analysis of Superalloy Disks With a Dual Grain Microstructure

Science.gov (United States)

Gayda, John; Kantzos, Pete

2006-01-01

Elastic-plastic finite element analyses of room temperature burst tests on four superalloy disks were conducted and reported in this paper. Two alloys, Rene 104 (General Electric Aircraft Engines) and Alloy 10 (Honeywell Engines & Systems), were studied. For both alloys an advanced dual microstructure disk, fine grain bore and coarse grain rim, were analyzed and compared with conventional disks with uniform microstructures, coarse grain for Rene 104 and fine grain for Alloy 10. The analysis and experimental data were in good agreement up to burst. At burst, the analysis underestimated the speed and growth of the Rene 104 disks, but overestimated the speed and growth of the Alloy 10 disks. Fractography revealed that the Alloy 10 disks displayed significant surface microcracking and coalescence in comparison to Rene 104 disks. This phenomenon may help explain the differences between the Alloy 10 disks and the Rene 104 disks, as well as the observed deviations between analytical and experimental data at burst.

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.)

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

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

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

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

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.)

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

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)

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

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)

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

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

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

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

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.

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.

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

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

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

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2009-11-15

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

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

International Nuclear Information System (INIS)

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

2009-01-01

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

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

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)

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

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)

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.

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

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.

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

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.

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

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.

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

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.)

Structure and properties of large-sized forged disk of alloy type KhN73MBTYu-VD(EhI 698-VD)

International Nuclear Information System (INIS)

Sudakov, V.S.

1994-01-01

Investigation results are presented for structure and mechanical properties of serial large-sized forged disk 1100 mm in diameter produced of alloy type EhI 9698-VD hand tested after standard heat treatment and isothermal ageing at operating temperature. Chemical composition studies have revealed no macroheterogeneity. In a central cross-section macrostructure is free of pores, inclusions, delaminating and variation in grain size. The metal of the disk possesses high values of long-term rupture strength and creep resistance at 650-700 deg C

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.

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

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

N18, powder metallurgy superalloy for disks: Development and applications

Energy Technology Data Exchange (ETDEWEB)

Guedou, J.Y.; Lautridou, J.C.; Honnorat, Y. (SNECMA, Evry (France). Materials and Processes Dept.)

1993-08-01

The preliminary industrial development of a powder metallurgy (PM) superalloy, designated N18, for disk applications has been completed. This alloy exhibits good overall mechanical properties after appropriate processing of the material. These properties have been measured on both isothermally forged and extruded billets, as well as on specimens cut from actual parts. The temperature capability of the alloy is about 700 C for long-term applications and approximately 750 C for short-term use because of microstructural instability. Further improvements in creep and crack propagation properties, without significant reduction in tensile strength, are possible through appropriate thermomechanical processing, which results in a large controlled grain size. Spin pit tests on subscale disks have confirmed that the N18 alloy has a higher resistance than PM Astrology and is therefore an excellent alloy for modern turbine disk applications.

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

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)

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

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.

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.

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

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

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.)

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

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

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

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

Disk-bend ductility tests for irradiated materials

International Nuclear Information System (INIS)

Klueh, R.L.; Braski, D.N.

1984-01-01

We modified the HEDL disk-bend test machine and are using it to qualitatively screen alloys that are susceptible to embrittlement caused by irradiation. Tests designed to understand the disk-bend test in relation to a uniaxial test are discussed. Selected results of tests of neutron-irradiated material are also presented

The Mechanical Properties of Candidate Superalloys for a Hybrid Turbine Disk

Science.gov (United States)

Gabb, Timothy P.; MacKay, Rebecca A.; Draper, Susan L.; Sudbrack, Chantal K.; Nathal, Michael V.

2013-01-01

The mechanical properties of several cast blade superalloys and one powder metallurgy disk superalloy were assessed for potential use in a dual alloy hybrid disk concept of joined dissimilar bore and web materials. Grain size was varied for each superalloy class. Tensile, creep, fatigue, and notch fatigue tests were performed at 704 to 815 degC. Typical microstructures and failure modes were determined. Preferred materials were then selected for future study as the bore and rim alloys in this hybrid disk concept. Powder metallurgy superalloy LSHR at 15 micron grain size and single crystal superalloy LDS-1101+Hf were selected for further study, and future work is recommended to develop the hybrid disk concept.

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.)

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

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

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

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.

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

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.

Bend testing for miniature disks

International Nuclear Information System (INIS)

Huang, F.H.; Hamilton, M.L.; Wire, G.L.

1982-01-01

A bend test was developed to obtain ductility measurements on a large number of alloy variants being irradiated in the form of miniature disks. Experimental results were shown to be in agreement with a theoretical analysis of the bend configuration. Disk specimens fabricated from the unstrained grip ends of previously tested tensile specimens were used for calibration purposes; bend ductilities and tensile ductilities were in good agreement. The criterion for estimating ductility was judged acceptable for screening purposes

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.

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.)

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

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

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

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

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.

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.

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)

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

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.

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.

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

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

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)

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.

Dwell Notch Low Cycle Fatigue Behavior of a Powder Metallurgy Nickel Disk Alloy

Science.gov (United States)

Telesman, J.; Gabb, T. P.; Yamada, Y.; Ghosn, L. J.; Jayaraman, N.

2012-01-01

A study was conducted to determine the processes which govern dwell notch low cycle fatigue (NLCF) behavior of a powder metallurgy (P/M) ME3 disk superalloy. The emphasis was placed on the environmentally driven mechanisms which may embrittle the highly stressed notch surface regions and reduce NLCF life. In conjunction with the environmentally driven notch surface degradation processes, the visco-plastic driven mechanisms which can significantly change the notch root stresses were also considered. Dwell notch low cycle fatigue testing was performed in air and vacuum on a ME3 P/M disk alloy specimens heat treated using either a fast or a slow cooling rate from the solutioning treatment. It was shown that dwells at the minimum stress typically produced a greater life debit than the dwells applied at the maximum stress, especially for the slow cooled heat treatment. Two different environmentally driven failure mechanisms were identified as the root cause of early crack initiation in the min dwell tests. Both of these failure mechanisms produced mostly a transgranular crack initiation failure mode and yet still resulted in low NLCF fatigue lives. The lack of stress relaxation during the min dwell tests produced higher notch root stresses which caused early crack initiation and premature failure when combined with the environmentally driven surface degradation mechanisms. The importance of environmental degradation mechanisms was further highlighted by vacuum dwell NLCF tests which resulted in considerably longer NLCF lives, especially for the min dwell tests.

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

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)

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)

Fabrication of Turbine Disk Materials by Additive Manufacturing

Science.gov (United States)

Sudbrack, Chantal; Bean, Quincy A.; Cooper, Ken; Carter, Robert; Semiatin, S. Lee; Gabb, Tim

2014-01-01

Precipitation-strengthened, nickel-based superalloys are widely used in the aerospace and energy industries due to their excellent environmental resistance and outstanding mechanical properties under extreme conditions. Powder-bed additive manufacturing (AM) technologies offer the potential to revolutionize the processing of superalloy turbine components by eliminating the need for extensive inventory or expensive legacy tooling. Like selective laser melting (SLM), electron beam melting (EBM) constructs three-dimensional dense components layer-by-layer by melting and solidification of atomized, pre-alloyed powder feedstock within 50-200 micron layers. While SLM has been more widely used for AM of nickel alloys like 718, EBM offers several distinct advantages, such as less retained residual stress, lower risk of contamination, and faster build rates with multiple-electron-beam configurations. These advantages are particularly attractive for turbine disks, for which excessive residual stress and contamination can shorten disk life during high-temperature operation. In this presentation, we will discuss the feasibility of fabricating disk superalloy components using EBM AM. Originally developed using powder metallurgy forging processing, disk superalloys contain a higher refractory content and precipitate volume fraction than alloy 718, thus making them more prone to thermal cracking during AM. This and other challenges to produce homogeneous builds with desired properties will be presented. In particular, the quality of lab-scale samples fabricated via a design of experiments, in which the beam current, build temperature, and beam velocity were varied, will be summarized. The relationship between processing parameters, microstructure, grain orientation, and mechanical response will be discussed.

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)

Microstructural evolution and thermophysical property evaluation of Th-U alloys

International Nuclear Information System (INIS)

Das, Santanu; Kaity, Santu; Bannerjee, Joydipto; Kumar, Raj; Roy, S.B.; Chaudhari, G.P.; Daniel, B.S.S.

2015-01-01

Thorium-uranium alloy fuel has not received much research attention mainly because of easy availability of uranium and military incentive offered by U-Pu cycle. Moreover, (i) lack of a consistent systematic effort to develop the alloys and define the limitations of these fuels, (ii) dearth of initiatives to define its microstructures that can result from composition and fabrication variables are prime reasons for this system not having witnessed much developmental research endeavour. Hence, it seems prudent to explore few compositions selected from thorium-uranium phase diagram keeping two primary objectives in view viz. (i) establishing its microstructural features and to study the variations in those, if any, brought about by processing variables etc. and (ii) to assess few thermal properties relevant to fuel applications. This experimental work aims at addressing gap in research on thorium-uranium alloys. Selected compositions of thorium-uranium alloy have been taken for microstructural study and evaluation of thermophysical properties. Based on the microstructural features and thermophysical property evaluation it is seen that high thorium Th-U alloys have appreciable thermal conductivity and low thermal expansion coefficient. It can reasonably be concluded that high thorium Th-U alloy can be used for possible nuclear fuel application in reactors provided other factors (e.g. reactor physics, post irradiation examinations etc.) are also seen to be favourable. (author)

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)

Ceramic Inclusions In Powder Metallurgy Disk Alloys: Characterization and Modeling

Science.gov (United States)

Bonacuse, Pete; Kantzos, Pete; Telesman, Jack

2002-01-01

Powder metallurgy alloys are increasingly used in gas turbine engines, especially as the material chosen for turbine disks. Although powder metallurgy materials have many advantages over conventionally cast and wrought alloys (higher strength, higher temperature capability, etc.), they suffer from the rare occurrence of ceramic defects (inclusions) that arise from the powder atomization process. These inclusions can have potentially large detrimental effect on the durability of individual components. An inclusion in a high stress location can act as a site for premature crack initiation and thereby considerably reduce the fatigue life. Because these inclusions are exceedingly rare, they usually don't reveal themselves in the process of characterizing the material for a particular application (the cumulative volume of the test bars in a fatigue life characterization is typically on the order of a single actual component). Ceramic inclusions have, however, been found to be the root cause of a number of catastrophic engine failures. To investigate the effect of these inclusions in detail, we have undertaken a study where a known population of ceramic particles, whose composition and morphology are designed to mimic the 'natural' inclusions, are added to the precursor powder. Surface connected inclusions have been found to have a particularly large detrimental effect on fatigue life, therefore the volume of ceramic 'seeds' added is calculated to ensure that a minimum number will occur on the surface of the fatigue test bars. Because the ceramic inclusions are irregularly shaped and have a tendency to break up in the process of extrusion and forging, a method of calculating the probability of occurrence and expected intercepted surface and embedded cross-sectional areas were needed. We have developed a Monte Carlo simulation to determine the distributions of these parameters and have verified the simulated results with observations of ceramic inclusions found in macro

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

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.

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

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)

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.

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

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

Microstructural Evaluations of Baseline HSR/EPM Disk Alloys

Science.gov (United States)

Gabb, Timothy P.; Garg, Anita; Ellis, David L.

2004-01-01

Six alloys representing two classes of powder metallurgy nickel-based superalloys were examined by transmission electron microscopy (TEM) and phase extraction. Alloys KM4, CH98, IN-100 and 456 are based on a Ni-18Co-12Cr composition while alloys Rene' 88 DT and SR 3 have lower Al and Co and higher Cr contents. The lambda size distributions were determined from quantitative image analysis of the TEM images. The volume fraction of lambda and carbides and the composition of the phases were determined by a combination of phase extraction and TEM. The results showed many similarities in lambda size distributions, grain boundary serrations, and grain boundary carbide frequencies between alloys KM4, CH98, 456, Rene' 88 DT and SR 3 when heat treated to give an approximate grain size of ASTM 6. The density of grain boundary carbides in KM4 was shown to substantially increase as the grain size increased. IN-100 and 456 subjected to a serration cooling heat treatment had much more complex lambda size distributions with very large intergranular and intragranular secondary lambda as well as finer than average cooling and aging lambda. The grain boundary carbides in IN-100 were similar to the other alloys, but 456 given the serration cooling heat treatment had a more variable density of grain boundary carbides. Examination of the phases extracted from the matrix showed that there were significant differences in the phase chemistries and elemental partitioning ratios between the various alloys.

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)

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

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

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.

Evolution of microstructure of U-Mo alloys in as cast and sintered forms

International Nuclear Information System (INIS)

Sinha, V.P.; Hegde, P.V.; Prasad, G.J.; Kamath, H.S.; Dey, G.K.

2009-01-01

Over the years U 3 Si 2 compound dispersed in aluminium matrix has been successfully used as potential Low Enriched Uranium (LEU 235 ) base dispersion fuel in new research and test reactors and also for converting High Enriched Uranium (HEU > 85% U 235 ) cores to LEU in most of the existing research and test reactors. The maximum density achievable with U 3 Si 2 -AI dispersion fuel is around 4.8 g U cm -3 . To achieve a uranium density of 8.0 to 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 Metallic Fuels Division, R and D efforts are on to develop these high density uranium alloys. Molybdenum plays a crucial role in metastabilising the γ-phase of uranium at room temperature which is very much evident when we see the microstructures of different U-Mo alloys with varying molybdenum concentration as solute atom. The paper describes the role of molybdenum in imparting metastability in U-Mo alloys from their microstructures in as cast and sintered forms. The paper also covers the role of tailored microstructure in U-Mo alloy for the purpose of hydriding and dehydriding treatment to generate alloy powders. (author)

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.)

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 ...

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

The hydrolysis of thorium dicarbide and of mixed uranium-thorium dicarbides; L'hydrolyse du dicarbure de thorium et des dicarbures mixtes d'uranium et de thorium

Energy Technology Data Exchange (ETDEWEB)

Del Litto, B [Commissariat a l' Energie Atomique, Grenoble (France). Centre d' Etudes Nucleaires

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) [French] L'hydrolyse du dicarbure de thorium conduit a la formation d'un melange complexe d'hydrures de carbone gazeux et condenses. La temperature entre 25 et 100 deg. C n'a pas d'influence sur la nature ef la composition de la phase gazeuse. Par contre la cinetique en depend fortement. En milieu chlorhydrique, on observe un enrichissement en hydrogene du melange gazeux. Au contraire, en milieu oxydant il se produit une diminution du taux d'hydrogene et une augmentation tres nette du taux d'acetylene. L'ensemble des resultats obtenus peut etre interprete d'une maniere

Uranium spectra in the ICP

Energy Technology Data Exchange (ETDEWEB)

Ghazi, A.A.; Qamar, S.; Atta, M.A. (Khan (A.Q.) Research Labs., Rawalpindi (Pakistan))

1994-05-01

Uranium spectra have been studied by inductively coupled plasma atomic emission spectroscopy (ICP-AES). In total, 8361 uranium lines were observed in the wavelength range of 235-500 nm. This article is an electronic publication in Spectrochimica Acta Electronica (SAE), the electronic section of Spectrochimica Acta Part B (SAB). The hard copy text is accompanied by a disk with data files and test files for an IBM-compatible computer. The main article discusses the scientific aspects of the subject and explains the purpose of the data files. (Author).

Uranium spectra in the ICP

International Nuclear Information System (INIS)

Ghazi, A.A.; Qamar, S.; Atta, M.A.

1994-01-01

Uranium spectra have been studied by inductively coupled plasma atomic emission spectroscopy (ICP-AES). In total, 8361 uranium lines were observed in the wavelength range of 235-500 nm. This article is an electronic publication in Spectrochimica Acta Electronica (SAE), the electronic section of Spectrochimica Acta Part B (SAB). The hard copy text is accompanied by a disk with data files and test files for an IBM-compatible computer. The main article discusses the scientific aspects of the subject and explains the purpose of the data files. (Author)

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)

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)

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)

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

MIT miniaturized disk bend test

International Nuclear Information System (INIS)

Harling, O.K.; Lee, M.; Sohn, D.S.; Kohse, G.; Lau, C.W.

1983-01-01

A miniaturized disk bend test (MDBT) using transmission electron microscopy specimens for the determination of various mechanical properties is being developed at MIT. Recent progress in obtaining strengths and ductilities of highly irradiated metal alloys is reviewed. Other mechanical properties can also be obtained using the MDBT approach. Progress in fatigue testing and in determination of the ductile-to-brittle transition temperature is reviewed briefly. 11 figures

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)

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

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

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

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

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.)

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)

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

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

Development of very-high-density low-enriched-uranium fuels

International Nuclear Information System (INIS)

Snelgrove, J.L.; Hofman, G.L.; Meyer, M.K.; Trybus, C.L.; Wiencek, T.C.

1997-01-01

Following a hiatus of several years and following its successful development and qualification of 4.8 g U cm -3 U 3 Si 2 -Al dispersion fuel for application with low-enriched uranium in research and test reactors, the US Reduced Enrichment for Research and Test Reactors program has embarked on the development of even-higher-density fuels. Our goal is to achieve uranium densities of 8-9 g cm -3 in aluminum-based dispersion fuels. Achieving this goal will require the use of high-density, γ-stabilized uranium alloy powders in conjunction with the most-advanced fuel fabrication techniques. Key issues being addressed are the reaction of the fuel alloys with aluminum and the irradiation behavior of the fuel alloys and any reaction products. Test irradiations of candidate fuels in very-small (micro) plates are scheduled to begin in the Advanced Test Reactor during June, 1997. Initial results are expected to be available in early 1998. We are performing out-of-reactor studies on the phase structure of the candidate alloys on diffusion of the matrix material into the aluminum. In addition, we are modifying our current dispersion fuel irradiation behavior model to accommodate the new fuels. Several international partners are participating in various phases of this work. (orig.)

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.

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

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.

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)

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)

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

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

Preparation of uranium electrodeposited target in aqueous system

International Nuclear Information System (INIS)

Chen Qiping; Li Yougen; Zhong Wenbin

2006-03-01

The main factors affecting uranium electrodeposition were tested and discussed. In the primary experiment about preparation of uranium isotopic target by electrodeposition, a stainless steel disk has been chosen as the target material, the electrolytic bath is comprised of UO 2 (NO 3 ) 2 and (NH 4 ) 2 C 2 O 4 , which has been adjusted to a pH of 2-3. Composition of the lost electrolytic bath was analysed by spectrophotometer. The thickness of resulting film is about 8-10 mg/cm 2 , the target having a thin, continuous, uniform layer of uranium, and its electrodeposited rate is more than 80%. (authors)

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)

Determination of ultratrace concentrations of uranium and thorium in natural waters by x-ray fluorescence

International Nuclear Information System (INIS)

Stewart, J.H. Jr.; Brooksbank, R.D.

1981-01-01

An x-ray fluorescence method for the simultaneous determination of uranium and thorium at the less than 1 ppM level in natural waters is described. Uranium and thorium are coprecipitated with an internal standard, yttrium, and incorporated into an iron-aluminum hydroxide carrier. The hydroxide precipitate is filtered, and the filter disk is analyzed by the energy-dispersive x-ray fluorescence technique. Matrix interferences caused by the presence of unpredictable anions and cations are compensated for by the internal standard. The U/Y and Th/Y ratios are linear over the 5- to 100-μg range of interest, and the detection limit of each element on the filter disk is 2 μg (3 sigma). Relative standard deviation was 17% at the 15-μg and 4% at the 100-μg level for thorium and 11% at the 11-μg and 2% at the 100-μg level for uranium. Analysis of spiked solutions showed a recovery of 19.6 +- 0.3 μg for uranium and 19.8 +- 0.3 μg for thorium at the 20-μg level, and the normal lower reporting limit is 5 μg. Fifty disks can be routinely measured during a normal working day

The participation of the Experimental Design Factory of the Uranium Industry of Czechoslovakia in the design of a tunneling machine with disk bits

Energy Technology Data Exchange (ETDEWEB)

Kastner, P

1983-01-01

A tunneling machine, two prototypes of which were designed and built jointly on the basis of scientific and technical cooperation between the Experimental Design Factory of the Uranium Industry of Czechoslovakia and the VEB-Schachtbau enterprise (East Germany), is described. The experimental design operations were conducted under the methodological leadership of the Mine Construction in the Uranium Industry (Czechoslovakia) enterprise. The experimental design factory developed a general design system for the machine and its individual subassemblies. The detailed technical documentation for the machine units was developed by both enterprises. Each enterprise made two complexes of specific units and spare parts. The prototypes were assembled in both countries with the technical assistance of the producer enterprise of the appropriate subassembly. Industrial tests were conducted by each enterprise independently with technical assistance and delivery of spare parts on the part of the producer enterprise. A machine under the title of VM 24-27 was used to drill more than 2,300 meters of water supply tunnel in East Germany in 1982 and a machine called the RS 24-27 (29) was used in Prague in the same year to drill approximately 1,400 meters of cable collectors. The machine is designed for the passage of rounded mine drifts with a diameter of 2.4 to 2.7 (2.9) meters) to the full cross section in stable rocks. Its overall length is 32.5 meters, while the total weight is 85 tons. The shift productivity was 9.55 meters. Since 1979 the Mining Construction in the Uranium Industry and the Experimental Design Plant of the Uranium Industry Enterprises of Czechoslovakia have supplied disk bits for the TVM Demag tunnel drilling machines (West Germany) and RS 24-27 and the HG 210 Wirth (West Germany) cross cut drills.

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

The MIT miniaturized disk bend test

International Nuclear Information System (INIS)

Harling, O.K.; Lee, M.; Sohn, D.S.; Kohse, G.; Lau, C.W.

1983-01-01

A miniaturized disk bend test (MDBT) using transmission electron microscopy specimens for the determination of various mechanical properties is being developed at MIT. Recent progress in obtaining strengths and ductilities of highly irradiated metal alloys is reviewed. Possibilities exist for applying the MDBT approach to the determination of other mechanical properties. Progress in fatigue testing and in determination of the ductile-to-brittle transition temperature is reviewed briefly

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

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

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

Progress toward uranium scrap recycling via electron beam cold hearth refining

International Nuclear Information System (INIS)

McKoon, R.H.

1994-01-01

A 250 kW electron beam cold hearth refining (EBCHR) melt furnace at Lawrence Livermore National Laboratory (LLNL) has been in operation for over a year producing 5.5 in.-diameter ingots of various uranium alloys. Production of in-specification uranium-6%-niobium (U-6Nb) alloy ingots has been demonstrated using Virgin feedstock. A vibratory scrap feeder has been installed on the system and the ability to recycle chopped U-6Nb scrap has been established. A preliminary comparison of vacuum arc remelted (VAR) and electron beam (EB) melted product is presented

Progress toward uranium scrap recycling via Electron Cold Hearth Refining (EBCHR)

International Nuclear Information System (INIS)

McKoon, R.H.

1994-01-01

A 250 kW electron beam cold hearth refining (EBCHR) melt furnace at Lawrence Livermore National Laboratory (LLNL) has been in operation for over a year producing 5.5 in.-diameter ingots of various uranium alloys. Production of in-specification uranium-6% - niobium (U-6Nb) alloy ingots has been demonstrated using virgin feedstock. A vibratory scrap feeder has been installed on the system and the ability to recycle chopped U-6Nb scrap has been established. A preliminary comparison of vacuum arc remelted (VAR) and electron beam (EB) melted product is presented

Simulation of uranium aluminide dissolution in a continuous aluminum dissolver system

International Nuclear Information System (INIS)

Evans, D.R.; Farman, R.F.; Christian, J.D.

1990-01-01

This paper reports on the Idaho Chemical Processing Plant (ICPP) which recovers highly-enriched uranium (uranium that contains at least 20 atom percent 235 U) from spent nuclear reactor fuel by dissolution of the fuel elements and extraction of the uranium from the aqueous dissolver product. Because the uranium is highly-enriched, consideration must be given to whether a critical mass can form at any stage of the process. In particular, suspended 235 U-containing particles are of special concern, due to their high density (6.8 g/cm 3 ) and due to the fact that they can settle into geometrically unfavorable configurations when not adequately mixed. A portion of the spent fuel is aluminum-alloy-clad uranium aluminide (UAl 3 ) particles, which dissolve more slowly than the cladding. As the aluminum alloy cladding dissolves in mercury-catalyzed nitric acid, UAl 3 is released. Under standard operating conditions, the UAl 3 dissolves rapidly enough to preclude the possibility of forming a critical mass anywhere in the system. However, postulated worst-case abnormal operating conditions retard uranium aluminide dissolution, and thus require evaluation. To establish safety limits for operating parameters, a computerized simulation model of uranium aluminide dissolution in the aluminum fuel dissolver system was developed

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

International Nuclear Information System (INIS)

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

1986-09-01

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

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)

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)

Preparation of uranium electrodeposited target in aqueous system

Energy Technology Data Exchange (ETDEWEB)

Qiping, Chen; Yougen, Li; Wenbin, Zhong [China Academy of Engineering Physics, Mianyang (China). Inst. of Nuclear Physics and Chemistry

2006-03-15

The main factors affecting uranium electrodeposition were tested and discussed. In the primary experiment about preparation of uranium isotopic target by electrodeposition, a stainless steel disk has been chosen as the target material, the electrolytic bath is comprised of UO{sub 2}(NO{sub 3}){sub 2} and (NH{sub 4}){sub 2}C{sub 2}O{sub 4}, which has been adjusted to a pH of 2-3. Composition of the lost electrolytic bath was analysed by spectrophotometer. The thickness of resulting film is about 8-10 mg/cm{sup 2}, the target having a thin, continuous, uniform layer of uranium, and its electrodeposited rate is more than 80%. (authors)

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

Progress on the IPNS Enriched Uranium Booster Target

International Nuclear Information System (INIS)

Knox, A.E.; Carpenter, J.M.; Bailey, J.L.

1986-09-01

We describe the Enriched Uranium Booster Target designed for use in Argonne's Intense Pulsed Neutron Source. This report contains a general description of the system, and descriptions of the thermal-hydraulic and loss-of-coolant accident analyses, of the neutronic, criticality and power density calculations, of the assessment of radiation and thermal cycling growth, and of the disk fabrication methods. We also describe the calculations of radionuclide buildup and the related hazards analysis and our calculations of the temperature and stress profiles in the disks, and briefly allude to considerations of security and safeguards

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

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.

Formation of uranium based nanoparticles via gamma-irradiation

Energy Technology Data Exchange (ETDEWEB)

Nenoff, Tina M., E-mail: tmnenof@sandia.gov [Nanoscale Sciences Department, Sandia National Laboratories, P.O. Box 5800, MS-1415, Albuquerque, NM 87185 (United States); Ferriera, Summer R. [Nanoscale Sciences Department, Sandia National Laboratories, P.O. Box 5800, MS-1415, Albuquerque, NM 87185 (United States); Huang, Jianyu [Center for Integrated Nanotechnology, Sandia National Laboratories, P.O. Box 5800, MS-1315, Albuquerque, NM 87185 (United States); Hanson, Donald J. [Department of Hot Cells and Gamma Facilities, Sandia National Laboratories, P.O. Box 5800, MS-1143, Albuquerque, NM 87185 (United States)

2013-11-15

Graphical abstract: TEM image of d-U nanoparticles formed in aqueous solution by gamma irradiation. Display Omitted -- Highlights: •d-U nanoparticles were grown in solution by gamma irradiation. •The reaction solution does not exceed 25 °C (room temperature). •Only after multiday exposure to air is there evidence of oxidation of the d-U nanoparticles. •Evidence of d-U alloy nanoparticle formation confirmed by TEM/energy-dispersive X-ray (EDS) analysis. -- Abstract: The ability to fabricate nuclear fuels at low temperatures allows for the production of complex Uranium metal and alloys with minimum volatility of alloy components in the process. Gamma irradiation is a valuable method for the synthesis of a wide range of metal-based nanoparticles. We report on the synthesis via room temperature radiolysis and characterization of uranium (depleted, d-U) metal and uranium–lathanide (d-ULn, Ln = lanthanide surrogates) alloy nanoparticles from aqueous acidic salt solutions. The lanthanide surrogates chosen include La and Eu due to their similarity in ionic size and charge in solution. Detailed characterization results including UV–vis, TEM/HR-TEM, and single particle EDX (elemental analyses) are presented for the room temperature formed nanoparticle products.

Formation of uranium based nanoparticles via gamma-irradiation

International Nuclear Information System (INIS)

Nenoff, Tina M.; Ferriera, Summer R.; Huang, Jianyu; Hanson, Donald J.

2013-01-01

Graphical abstract: TEM image of d-U nanoparticles formed in aqueous solution by gamma irradiation. Display Omitted -- Highlights: •d-U nanoparticles were grown in solution by gamma irradiation. •The reaction solution does not exceed 25 °C (room temperature). •Only after multiday exposure to air is there evidence of oxidation of the d-U nanoparticles. •Evidence of d-U alloy nanoparticle formation confirmed by TEM/energy-dispersive X-ray (EDS) analysis. -- Abstract: The ability to fabricate nuclear fuels at low temperatures allows for the production of complex Uranium metal and alloys with minimum volatility of alloy components in the process. Gamma irradiation is a valuable method for the synthesis of a wide range of metal-based nanoparticles. We report on the synthesis via room temperature radiolysis and characterization of uranium (depleted, d-U) metal and uranium–lathanide (d-ULn, Ln = lanthanide surrogates) alloy nanoparticles from aqueous acidic salt solutions. The lanthanide surrogates chosen include La and Eu due to their similarity in ionic size and charge in solution. Detailed characterization results including UV–vis, TEM/HR-TEM, and single particle EDX (elemental analyses) are presented for the room temperature formed nanoparticle products

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)

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

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.

Influence of Temperature to Thermal Properties of U-Zr Alloy With The Zr Content Variation

International Nuclear Information System (INIS)

Aslina-Br-Ginting; Masrukan; M-Husna-Al-Hasa

2007-01-01

Have been done thermal of characteristic covering heat stability, heat capacities, enthalpy and also phase changes from uranium, zirkonium and U-Zr alloy with the Zr content variation of Zr 2 %, 6 %, 10% and 14% weight. Change of the temperature and composition anticipated will cause the characteristic of thermal to uranium metal, zirkonium and also U-Zr alloy. Therefore at this research was conducted using analysis influence of temperature to thermal of characteristic of uranium, zirkonium and U-Zr alloy with the Zr content variation by using DTA and DSC. Result of analysis indicate that the uranium metal at temperature 662 o C stable in phase α. Above at temperature, uranium metal experience of the phase change indicated by formed the thermochemical reaction as much 3 endothermic peak. At temperature 667.16 o C, happened by the phase change of α become the phase β with the enthalpy 2,3034 cal/g, at temperature 773.05 o C happened by the phase change β becoming phase γ 2,8725 cal/g and also at temperature 1125.26 the o C uranium metal experience the phenomenon become to melt with the enthalpy 2,1316 cal/g. (author)

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.)

Thermodynamic activity measurements of U-Zr alloys by Knudsen effusion mass spectrometry

International Nuclear Information System (INIS)

Kanno, Masayoshi; Yamawaki, Michio; Koyama, Tadafumi; Morioka, Nobuo

1988-01-01

Vaporization of a series of U-Zr alloys, a fundamental subsystem of the promising metallic fuel U-Pu-Zr, was studied by using a tantalum Knudsen cell coupled with a mass spectrometer in the temperature range 1700-2060 K. Thermodynamic activities partial molar Gibbs free energies and integral molar Gibbs free energies of mixing were calculated from the partial vapor pressures of uranium over these alloys. The activities of uranium exhibit negative deviations from ideality, especially in the uranium-rich composition region. Both the solidus and liquidus lines for this system estimated from the activities show negative deviations from the tentative phase diagram previously reported. (orig.)

MEANS 2: Microstructure- and Micromechanism-Sensitive Property Models for Advanced Turbine Disk and Blade Systems

National Research Council Canada - National Science Library

Pollock, Tresa M; Mills, Michael J

2008-01-01

This effort has focused on verification and refinement of the mechanism transitions at intermediate temperatures in the disk alloy Rene 104, with the observation of microtwinning, continuous faulting...

Wear resistance of a pressable low-fusing ceramic opposed by dental alloys.

Science.gov (United States)

Faria, Adriana Cláudia Lapria; de Oliveira, André Almeida; Alves Gomes, Érica; Silveira Rodrigues, Renata Cristina; Faria Ribeiro, Ricardo

2014-04-01

Dental alloys have increasingly replaced by dental ceramics in dentistry because of aesthetics. As both dental alloys and ceramics can be present in the oral cavity, the evaluation of the wear resistance of ceramics opposed by dental alloys is important. The aim of the present study was to evaluate wear resistance of a pressable low-fusing ceramic opposed by dental alloys as well as the microhardness of the alloys and the possible correlation of wear and antagonist microhardness. Fifteen stylus tips samples of pressable low-fusing ceramic were obtained, polished and glazed. Samples were divided into three groups according to the disk of alloy/metal to be used as antagonist: Nickel-Chromium (Ni-Cr), Cobalt-Chromium (Co-Cr) and commercially pure titanium (cp Ti). Vickers microhardness of antagonist disks was evaluated before wear tests. Then, antagonist disks were sandblasted until surface roughness was adjusted to 0.75μm. Wear tests were performed at a speed of 60 cycles/min and distance of 10mm, in a total of 300,000 cycles. Before and after wear tests, samples were weighted and had their profile designed in an optical comparator to evaluate weight and height loss, respectively. Ni-Cr and cp Ti caused greater wear than Co-Cr, presenting greater weight (p=.009) and height (p=.002) loss. Cp Ti microhardness was lower than Ni-Cr and Co-Cr (pceramic presents different wear according to the dental alloy used as antagonist and the wear is not affected by antagonist microhardness. Copyright © 2013 Elsevier Ltd. All rights reserved.

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)

Comparison of the tensile bond strength of high-noble, noble, and base metal alloys bonded to enamel.

Science.gov (United States)

Sen, D; Nayir, E; Pamuk, S

2000-11-01

Although the bond strengths of various resin composite luting materials have been reported in the literature, the evaluation of these systems with various cast alloys of different compositions has not been completely clarified. To evaluate the tensile bond strength of sandblasted high-noble, noble, and base metal alloys bonded to etched enamel by 2 different bonding agents of different chemical composition: Panavia-Ex (BIS-GMA) and Super-Bond (4-META acrylic). Flat enamel surfaces were prepared on buccal surfaces of 60 extracted noncarious human incisors. Teeth were divided into 3 groups of 20 each. Twenty circular disks of 5 mm diameter were prepared for casting for each group. Group I was cast with a high-noble, group II with a noble, and group III with a base metal alloy. The surfaces of the disks were sandblasted with 250 microm Al(2)O(3). Ten disks of each group were bonded to exposed enamel surfaces with Super-Bond and 10 disks with Panavia-Ex as recommended by the manufacturer. The tensile bond strength was measured with an Instron universal testing machine with a crosshead speed of 0.5 mm/min until failure occurred. Two-way ANOVA was used to evaluate the results. The differences in bond strengths of Super-Bond and Panavia-Ex with different alloys were not significant. The highest bond strengths were obtained in base metal alloys, followed by noble and high-noble alloys. These results were significant. Panavia-Ex and Super-Bond exhibited comparable tensile bond strengths. For both luting agents, the highest bond strengths were achieved with base metal alloys and the lowest with high-noble alloys.

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

Oxidation Behavior of Simudated Metallic U-Nb Alloys in Air

International Nuclear Information System (INIS)

Lee, Eun Pyo; Ju, June Sik; You, Gil Sung; Cho, Il Je; Kook, Dong Hak; Kim, Ho Dong

2004-01-01

In order to enhance an oxidation resistance of the pure uranium metal under air condition, a small quantity of niobium(Nb) which is known to mitigate metal oxidation is added into uranium metal as an alloying element. A simulated metallic uranium alloy, U-Nb has been fabricated and then oxidized in the range of 200 to under the environment of the pure oxygen gas. The oxidized quantity in terms of the weight gain (wt%) has been measured with the help of a thermogravimetric analyzer. The results show that the oxidation resistance of the U-Nb alloy is considerably enhanced in comparison with that of the pure uranium metal. It is revealed that the oxidation resistance of the former with the niobium content of 1, 2, 3, and 4 wt% is : 1) 1.61, 7.78, 11.76 and 20.14 times at the temperature of 200 .deg. C; 2) 1.45, 5.98, 10.08 and 11.15 times at 250 .deg. C; and 3) 1.33, 4.82, 8.87 and 6.84 times at 300 .deg. C higher than that of the latter, respectively. Besides, it is shown that the activation energy attributable to the oxidation is 17.13-21.92 kcal/mol.

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.

Behaviour of uranium under irradiation; Comportement de l'uranium sous irradiation

Energy Technology Data Exchange (ETDEWEB)

Adda, Y; Mustelier, J P; Quere, Y [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires; Commissariat a l' Energie Atomique, Fontenay-aux-Roses (France). Centre d' Etudes Nucleaires

1964-07-01

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

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

Effects of carbon and hafnium concentrations in wrought powder-metallurgy superalloys based on nasa 2b-11 alloy

International Nuclear Information System (INIS)

Miner, R.V. Jr.

1976-01-01

A candidate alloy for advanced-temperature turbine engine disks and four modifications of that alloy with various C and Hf concentrations were produced as cross-rolled disks from prealloyed powder that was hot isostatically compacted. The mechanical properties, microstructures, and phase relations of the alloys are discussed in terms of their C and Hf concentrations. A low-C and high-Hf modification of IIB-11 had the best balance of mechanical properties for service below about 750 C. Because of their finer grain sizes, none of the powder-metallurgy alloys produced had the high-temperature rupture strength of conventionally cast and wrought IIB-11. (Author)

Design Study for a Low-Enriched Uranium Core for the High Flux Isotope Reactor, Annual Report for FY 2008

Energy Technology Data Exchange (ETDEWEB)

Primm, Trent [ORNL; Chandler, David [ORNL; Ilas, Germina [ORNL; Miller, James Henry [ORNL; Sease, John D [ORNL; Jolly, Brian C [ORNL

2009-03-01

This report documents progress made during FY 2008 in studies of converting the High Flux Isotope Reactor (HFIR) from highly enriched uranium (HEU) fuel to low-enriched uranium (LEU) fuel. Conversion from HEU to LEU will require a change in fuel form from uranium oxide to a uranium-molybdenum alloy. With axial and radial grading of the fuel foil and an increase in reactor power to 100 MW, calculations indicate that the HFIR can be operated with LEU fuel with no degradation in reactor performance from the current level. Results of selected benchmark studies imply that calculations of LEU performance are accurate. Scoping experiments with various manufacturing methods for forming the LEU alloy profile are presented.

The substitution of nickel for cobalt in hot isostatically pressed powder metallurgy UDIMET 700 alloys

Science.gov (United States)

Harf, F. H.

1985-01-01

Nickel was substituted in various proportions for cobalt in a series of five hot-isostatically-pressed powder metallurgy alloys based on the UDIMET 700 composition. These alloys were given 5-step heat treatments appropriate for use in turbine engine disks. The resultant microstructures displayed three distinct sizes of gamma-prime particles in a gamma matrix. The higher cobalt-content alloys contained larger amounts of the finest gamma-prime particles, and had the lowest gamma-gamma-prime lattice mismatch. While all alloys had approximately the same tensile properties at 25 and 650 gamma C, the rupture lives at 650 and 760 C peaked in the alloys with cobalt contents between 12.7 and 4.3 pct. Minimum creep rates increased as cobalt contents were lowered, suggesting their correlation with the gamma-prime particle size distribution and the gamma-gamma-prime mismatch. It was also found that, on overaging at temperatures higher than suitable for turbine disk use, the high cobalt-content alloys were prone to sigma phase formation.

Microstructural and thermodynamic evaluation of as-cast U-rich U-Zr alloys

International Nuclear Information System (INIS)

Basak, Chandrabhanu; Prasad, G.J.; Kamath, H.S.

2009-01-01

The present study involves evaluation of microstructures and some basic properties of as-cast uranium rich U-Zr alloys; i.e. uranium alloys containing 2wt%, 5wt%, 7wt% and 10 wt% zirconium. Microstructural evaluation, both optical and SEM, with hardness values are reported. It was shown that a definite correlation exists between the microstructure and the hardness of the alloy. Lattice parameter and densities are determined with the help of XRD analysis. Also the phase transformation mechanism is proposed based on the microstructures and XRD analysis. Thermodynamic analysis coupled with the experimental observation reveals that the lamellar structure found in the as-cast U-rich U-Zr alloys originates from the monotectoid reaction (γ→β + γ'). As Zr concentration increases in the alloy the gamma phase can remain in the metastable state even at lower T. So, with increasing Zr content the monotectoid reaction takes place at lower temperature causing generation of finer lamellae. (author)

A Very High Uranium Density Fission Mo Target Suitable for LEU Using atomization Technology

Energy Technology Data Exchange (ETDEWEB)

Kim, C. K.; Kim, K. H.; Lee, Y. S.; Ryu, H. J.; Woo, Y. M.; Jang, S. J.; Park, J. M.; Choi, S. J. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2011-05-15

Currently HEU minimization efforts in fission Mo production are underway in connection with the global threat reduction policy. In order to convert HEU to LEU for the fission Mo target, higher uranium density material could be applied. The uranium aluminide targets used world widely for commercial {sup 99}Mo production are limited to 3.0 g-U/cc in uranium density of the target meat. A consideration of high uranium density using the uranium metal particles dispersion plate target is taken into account. The irradiation burnup of the fission Mo target are as low as 8 at.% and the irradiation period is shorter than 7 days. Pure uranium material has higher thermal conductivity than uranium compounds or alloys. It is considered that the degradation by irradiation would be almost negligible. In this study, using the computer code of the PLATE developed by ANL the irradiation behavior was estimated. Some considerations were taken into account to improve the irradiation performance further. It has been known that some alloying elements of Si, Cr, Fe, and Mo are beneficial for reducing the swelling by grain refinement. In the RERTR program recently the interaction problem could be solved by adding a small amount of Si to the aluminum matrix phase. The fabrication process and the separation process for the proposed atomized uranium particles dispersion target were reviewed

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)

Rotating disk atomization of Gd and Gd-Y for hydrogen liquefaction via magnetocaloric cooling

Energy Technology Data Exchange (ETDEWEB)

Slinger, Tyler [Iowa State Univ., Ames, IA (United States)

2016-12-17

In order to enable liquid hydrogen fuel cell technologies for vehicles the cost of hydrogen liquefaction should be lowered. The current method of hydrogen liquefaction is the Claude cycle that has a figure of merit (FOM) of 0.3-0.35. New magnetocaloric hydrogen liquefaction devices have been proposed with a FOM>0.5, which is a significant improvement. A significant hurdle to realizing these devices is the synthesis of spherical rare earth based alloy powders of 200μm in diameter. In this study a centrifugal atomization method that used a rotating disk with a rotating oil quench bath was developed to make gadolinium and gadolinium-yttrium spheres. The composition of the spherical powders included pure Gd and Gd_0.91Y_0.09. The effect of atomization parameters, such as superheat, melt properties, disk shape, disk speed, and melt system materials and design, were investigated on the size distribution and morphology of the resulting spheres. The carbon, nitrogen, and oxygen impurity levels also were analyzed and compared with the magnetic performance of the alloys. The magnetic properties of the charge material as well as the resulting powders were measured using a vibrating sample magnetometer. The saturation magnetization and Curie temperature were the target properties for the resulting spheres. These values were compared with measurements taken on the charge material in order to investigate the effect of atomization processing on the alloys.

Progress in developing very-high-density low-enriched-uranium fuels

International Nuclear Information System (INIS)

Snelgrove, J.L.; Hofman, G.L.; Meyer, M.K.; Hayes, S.L.; Wiencek, T.C.; Strain, R.V.

1999-01-01

Preliminary results from the postirradiation examinations of microplates irradiated in the RERTR-1 and -2 experiments in the ATR have shown several binary and ternary U-Mo alloys to be promising candidates for use in aluminum-based dispersion fuels with uranium densities up to 8 to 9 g/cm 3 . Ternary alloys of uranium, niobium, and zirconium performed poorly, however, both in terms of fuel/matrix reaction and fission-gas-bubble behavior, and have been dropped from further study. Since irradiation temperatures achieved in the present experiments (approximately 70 deg. C) are considerably lower than might be experienced in a high-performance reactor, a new experiment is being planned with beginning-of-cycle temperatures greater than 200 deg. C in 8-g U/cm 3 fuel. (author)

Microstructural characteristics of DU-xMo alloys with x = 7-12 wt%

International Nuclear Information System (INIS)

Burkes, Douglas E.; Hartmann, Thomas; Prabhakaran, Ramprashad; Jue, J.-F.

2009-01-01

Microstructural, phase, and impurity analyses of six depleted uranium-molybdenum alloys were obtained using optical metallography, X-ray diffraction, and carbon/nitrogen/oxygen determination. Uranium-molybdenum alloy foils are currently under investigation for the conversion of high-power research reactors using high-enriched uranium fuel to accommodate the use of low-enriched uranium fuel. Understanding basic microstructural behavior of these foils is an important consideration in determining the impact of fabrication processes and in anticipating performance of the foils in a reactor. Average grain diameter decreased with increasing molybdenum content. Lattice parameter decreased with increasing molybdenum content, and no significant degree of phase decomposition or crystallographic ordering was caused by processing and post-processing conditions employed in this study. Impurity concentration, specifically carbon, inhibited the degree of microstructural recrystallization but did not appear to impact other microstructural traits, such as γ-phase retention or lattice parameter.

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)

Disk

NARCIS (Netherlands)

P.A. Boncz (Peter); L. Liu (Lei); M. Tamer Özsu

2008-01-01

htmlabstractIn disk storage, data is recorded on planar, round and rotating surfaces (disks, discs, or platters). A disk drive is a peripheral device of a computer system, connected by some communication medium to a disk controller. The disk controller is a chip, typically connected to the CPU of

Disposal of waste computer hard disk drive: data destruction and resources recycling.

Science.gov (United States)

Yan, Guoqing; Xue, Mianqiang; Xu, Zhenming

2013-06-01

An increasing quantity of discarded computers is accompanied by a sharp increase in the number of hard disk drives to be eliminated. A waste hard disk drive is a special form of waste electrical and electronic equipment because it holds large amounts of information that is closely connected with its user. Therefore, the treatment of waste hard disk drives is an urgent issue in terms of data security, environmental protection and sustainable development. In the present study the degaussing method was adopted to destroy the residual data on the waste hard disk drives and the housing of the disks was used as an example to explore the coating removal process, which is the most important pretreatment for aluminium alloy recycling. The key operation points of the degaussing determined were: (1) keep the platter plate parallel with the magnetic field direction; and (2) the enlargement of magnetic field intensity B and action time t can lead to a significant upgrade in the degaussing effect. The coating removal experiment indicated that heating the waste hard disk drives housing at a temperature of 400 °C for 24 min was the optimum condition. A novel integrated technique for the treatment of waste hard disk drives is proposed herein. This technique offers the possibility of destroying residual data, recycling the recovered resources and disposing of the disks in an environmentally friendly manner.

U-Zr alloy: XPS and TEM study of surface passivation

Science.gov (United States)

Paukov, M.; Tkach, I.; Huber, F.; Gouder, T.; Cieslar, M.; Drozdenko, D.; Minarik, P.; Havela, L.

2018-05-01

Surface reactivity of Uranium metal is an important factor limiting its practical applications. Bcc alloys of U with various transition metals are much less reactive than pure Uranium. So as to specify the mechanism of surface protection, we have been studying the U-20 at.% Zr alloy by photoelectron spectroscopy and transmission electron microscopy. The surface was studied in as-obtained state, in various stages of surface cleaning, and during an isochronal annealing cycle. The analysis based on U-4f, Zr-3p, and O-1 s spectra shows that a Zr-rich phase segregates at the surface at temperatures exceeding 550 K, which provides a self-assembled coating. The comparison of oxygen exposure of the stoichiometric and coated surfaces shows that the coating is efficiently preventing the oxidation of uranium even at elevated temperatures. The coating can be associated with the UZr2+x phase. TEM study indicated that the coating is about 20 nm thick. For the clean state, the U-4f core-level lines of the bcc alloy are practically identical to those of α-U, revealing similar delocalization of the 5f electronic states.

Heat-induced redistribution of surface oxide in uranium

International Nuclear Information System (INIS)

Swissa, E.; Shamir, N.; Bloch, J.; Mintz, M.H.; Israel Atomic Energy Commission, Beersheba. Nuclear Research Center-Negev)

1990-01-01

The redistribution of oxygen and uranium metal at the vicinity of the metal-oxide interface of native and grown oxides due to vacuum thermal annealing was studied for uranium and uranium-chromium alloy using Auger depth profiling and metallographic techniques. It was found that uranium metal is segregating out through the uranium oxide layer for annealing temperatures above 450deg C. At the same time the oxide is redistributed in the metal below the oxide-metal interface in a diffusion like process. By applying a diffusion equation of a finite source, the diffusion coefficients for the process were obtained from the oxygen depth profiles measured for different annealing times. An Arrhenius like behavior was found for the diffusion coefficient between 400 and 800deg C. The activation energy obtained was E a =15.4±1.9 kcal/mole and the pre-exponential factor, D 0 =1.1x10 -8 cm 2 /s. An internal oxidation mechanism is proposed to explain the results. (orig.)

Heat-induced redistribution of surface oxide in uranium

Science.gov (United States)

Swissa, Eli; Shamir, Noah; Mintz, Moshe H.; Bloch, Joseph

1990-09-01

The redistribution of oxygen and uranium metal at the vicinity of the metal-oxide interface of native and grown oxides due to vacuum thermal annealing was studied for uranium and uranium-chromium alloy using Auger depth profiling and metallographic techniques. It was found that uranium metal is segregating out through the uranium oxide layer for annealing temperatures above 450°C. At the same time the oxide is redistributed in the metal below the oxide-metal interface in a diffusion like process. By applying a diffusion equation of a finite source, the diffusion coefficients for the process were obtained from the oxygen depth profiles measured for different annealing times. An Arrhenius like behavior was found for the diffusion coefficient between 400 and 800°C. The activation energy obtained was Ea = 15.4 ± 1.9 kcal/mole and the pre-exponential factor, D0 = 1.1 × 10 -8cm2/ s. An internal oxidation mechanism is proposed to explain the results.

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

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

Development of 99Mo isotope production targets employing uranium metal foils

International Nuclear Information System (INIS)

Hofman, G.L.; Wiencek, T.C.; Wood, E.L.; Snelgrove, J.L.

1997-01-01

The Reduced Enrichment Research and Test Reactor Program has continued its effort in the past 3 yr to develop use of low-enriched uranium (LEU) to produce the fission product 99 Mo. This work comprises both target and chemical processing development and demonstration. Two major target systems are now being used to produce 99 Mo with highly enriched uranium-one employing research reactor fuel technology (either uranium-aluminum alloy or uranium aluminide-aluminum dispersion) and the other using a thin deposit of UO 2 on the inside of a stainless steel (SST) tube. This paper summarizes progress in irradiation testing of targets based on LEU uranium metal foils. Several targets of this type have been irradiated in the Indonesian RSG-GAS reactor operating at 22.5 MW

Gamma stability and powder formation of UMo alloys

Energy Technology Data Exchange (ETDEWEB)

Oliveira, F.B.V.; Andrade, D.A.; Angelo, G.; Belchior Junior, A.; Torres, W.M.; Umbehaun, P.E., E-mail: wmtorres@ipen.br, E-mail: umbehaun@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil); Angelo, E., E-mail: eangelo@mackenzie.br [Universidade Presbiteriana Mackenzie, Sao Paulo, SP (Brazil). Grupo de Simulacao Numerica (GSN)

2015-07-01

A study of the hydrogen embrittlement as well as a research on the relation between gamma decomposition and powder formation of uranium molybdenum alloys were previously presented. In this study a comparison regarding the hypo-eutectoid and hyper-eutectoid molybdenum additions is presented. Gamma uranium molybdenum alloys have been considered as the fuel phase in plate type fuel elements for material and test reactors (MTR). Regarding their usage as a dispersion phase in aluminum matrix, it is necessary to convert the as cast structure into powder, and one of the techniques considered for this purpose is the hydration-dehydration (HDH). This paper shows that, under specific conditions of heating and cooling, γ-UMo fragmentation may occur with non-reactive or reactive mechanisms. Following the production of the alloys by induction melting, samples of the alloys were thermally treated under a constant flow of hydrogen. It was observed that, even without a massive hydration-dehydration process, the alloys fragmented under specific conditions of thermal treatment, during the thermal shock phase of the experiments. Also, there is a relation between absorption and the rate of gamma decomposition or the gamma phase stability of the alloy and this phenomenon can be related to the eutectoid transformation temperature. This study was carried out to search for a new method for the production of powders and for the evaluation of important physical parameter such as the eutectoid transformation temperature, as an alternative to the existing ones. (author)

Oxygen reduction reaction of Pt–In alloy: Combined theoretical and experimental investigations

International Nuclear Information System (INIS)

Pašti, Igor A.; Gavrilov, Nemanja M.; Baljozović, Miloš; Mitrić, Miodrag; Mentus, Slavko V.

2013-01-01

Graphical abstract: Upon DFT prediction of improved electrocatalytic activity of Pt–In alloys toward ORR, the alloy Pt-10 at% In was synthesized on glassy carbon disc, simultaneously with pure Pt reference catalyst. Improved catalytic activity of the alloy was evidenced by voltammetry on RDE in 0.1 mol dm −3 KOH solution. -- Highlights: •The adsorption of O atoms on Pt–In alloys model surfaces was investigated by DFT. •The improvement of catalytic activity toward ORR was predicted by DFT. •Pt-10 at% In alloy was synthesized on glassy carbon disk surface. •By voltammetry on RDE improvement of activity toward ORR was evidenced. -- Abstract: By means of the density functional theory (DFT) calculations, using the adsorption energy of oxygen on single crystal surfaces as criterion, it was predicted that the alloying of Pt with In should improve kinetics of oxygen reduction reaction (ORR). To prove this, the Pt–In alloy having nominal composition Pt 9 In was synthesized by heating H 2 PtCl 6 –InCl 3 mixture in hydrogen stream. The XRD characterization confirmed that Pt–In alloy was formed. The electrochemical measurements by rotating disk technique in alkaline 0.1 mol dm −3 KOH solution evidenced faster ORR kinetics for factor 2.6 relative to the one on pure platinum. This offers the possibility of searching for new ORR electrocatalysts by alloying platinum with p-elements

Study on uranium metallization yield of spent Pressurized Water Reactor fuels and oxidation behavior of fission products in uranium metals

International Nuclear Information System (INIS)

Choi, Ke Chon; Lee, Chang Heon; Kim, Won Ho

2003-01-01

Metallization yield of uranium oxide to uranium metal from lithium reduction process of spent Pressurized Water Reactor (PWR) fuels was measured using thermogravimetric analyzer. A reduced metal produced in the process was divided into a solid and a powder part, and each metallization yield was measured. Metallization yield of the solid part was 90.7∼95.9 wt%, and the powder being 77.8∼71.5 wt% individually. Oxidation behaviour of the quarternary alloy was investigated to take data on the thermal oxidation stability necessary for the study on dry storage of the reduced metal. At 600∼700 .deg. C, weight increments of allow of No, Ru, Rh and Pd was 0.40∼0.55 wt%. Phase change on the surface of the allow was started at 750 .deg. C. In particular, Mo was rapidly oxidized and then the alloy lost 0.76∼25.22 wt% in weight

Microstructural study on gamma phase stability in U-9 wt% Mo alloy system

International Nuclear Information System (INIS)

Saify, M.T.; Jha, S.K.; Hussain, M.M.; Singh, R.P.; Neogy, S.; Srivastava, D.; Dey, G.K.

2009-01-01

Uranium exists in three polymorphic forms viz., orthorhombic α phase - stable up to 667 deg C, tetragonal β phase - stable between 667 deg C and 771 deg C and bcc γ phase - stable above 771 deg C. When alloying of uranium is done, the alloying additions alter the temperature ranges over which the α, β and γ phases are stable. In addition, they frequently retard the rates at which phase transformations occur. As a result, a number of metastable phases can be obtained in uranium alloys. It has been well known among reactor designers that a pure uranium metal is not suitable for power reactor fuel mainly because of (i) phase changes occurring at lower temperatures and (ii) poor irradiation behavior of α phase. γ phase uranium alloys containing small amount of another metal to stabilize the γ-U solid solution provides good prospects in this respect. U-Mo alloy is one of the prospective materials for low enrichment uranium fuel with high U loading because a solid solution of Mo in the γ-U phase possesses acceptable irradiation and mechanical properties and is formed over a wide range of Mo concentration. In the present work vacuum induction melted and cast U-9 wt% Mo alloy was subjected to different thermo mechanical processing to investigate the stability of the γ phase. The as cast alloy was rolled at 550 deg C and then homogenized at 1000 deg C in the γ phase field for 24 hours followed by (i) water quenching and (ii) furnace cooling to generate two different starting conditions. Two of the water-quenched samples were aged at 500 deg C for 5 days and 14 days and one as-rolled sample was aged at 500 deg C for 5 days. The as-cast, as-rolled, homogenized and aged samples were subjected to optical microscopy and X-ray Diffraction (XRD) investigations. All the samples were also subjected to microhardness measurements. The as cast sample contained predominantly the gamma phase along with inclusions. After homogenizing the alloy at 1000 deg C and quenching in

Oxygen reduction of several gold alloys in 1-molar potassium hydroxide

Science.gov (United States)

Miller, R. O.

1975-01-01

With rotated disk-and-ring equipment, polarograms and other electrochemical measurements were made of oxygen reduction in 1-molar potassium hydroxide on an equiatomic gold-copper (Au-Cu) alloy and a Au-Cu alloy doped with either indium (In) or cobalt (Co) and on Au doped with either nickel (Ni) or platinum (Pt). The results were compared with those for pure Au and pure Pt. The two-electron reaction dominated on all Au alloys as it did on Au. The polarographic results at lower polarization potentials were compared, assuming exclusively a two-step reduction. A qualified ranking of cathodic electrocatalytic activity on the freshly polished reduced disks was indicated: anodized Au Au-Cu-In Au-Cu Au-Cu-Co is equivalent or equal to Au-Pt Au-Ni. Aging in distilled water improved the electrocatalytic efficiency of Au-Cu-Co, Au-Cu, and (to a lesser extent) Au-Cu-In.

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

Recovery of uranium from uranium and lanthanides in LiCl-KCl molten salt by electrowinning including Cd-Li anode

International Nuclear Information System (INIS)

Woo, Moon Shik; Kim, Eung Ho

2005-01-01

A trans-uranium (TRU) fuel should be manufactured and loaded in transmutation systems in order to transmute the long-lived TRU nuclides into short-lived ones. However, since all of the TRU nuclides are not completely transmuted in one cycle lifetime in transmutation systems, the spent TRU fuel has to be treated to recover the long-lived radionuclides or fuel matrix materials. One concept to manufacture TRU fuel for transmutation is to recover uranium from TRU and molten salt. If this type of fuel is adopted for transmutation, uranium could also be an objective material to be recovered and recycled. Since electrowinning is a promising technology to be employed for the recovery of uranium from fuel materials, some experimental work of electrowinning using anode of Cd-Li alloy was carried out in this study. The basic salt chosen was a mixture of LiCl-KCl which has an eutectic point at 357 .deg. C

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

Nuclear fuel cycle head-end enriched uranium purification and conversion into metal

International Nuclear Information System (INIS)

Bonini, A.; Cabrejas, J.; Lio, L. de; Dell'Occhio, L.; Devida, C.; Dupetit, G.; Falcon, M.; Gauna, A.; Gil, D.; Guzman, G.; Neuringer, P.; Pascale, A.; Stankevicius, A.

1998-01-01

The CNEA (Comision Nacional de Energia Atomica - Argentina) operated two facilities at the Ezeiza Atomic Center which supply purified enriched uranium employed in the production of nuclear fuels. At one of those facilities, the Triple Height Laboratory scraps from the production of MTR type fuel elements (mainly out of specification U 3 O 8 plates or powder) are purified to nuclear grade. The purification is accomplished by a solvent extraction process. The other facility, the Enriched Uranium Laboratory produces 90% enriched uranium metal to be used in Mo 99 production (originally the uranium was used for the manufacture of MTR fuel elements made of aluminium-uranium alloy). This laboratory also provided metallic uranium with a lower enrichment (20%) for a first uranium-silicon testing fuel element, and in the near future it is going to recommence 20% enriched uranium related activities in order to provide the metal for the silicon-based fuel elements production (according to the policy of enrichment reduction for MTR reactors). (author)

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.

Behavior of metallic uranium-fissium fuel in TREAT transient overpower tests

International Nuclear Information System (INIS)

Bauer, T.H.; Klickman, A.E.; Lo, R.K.; Rhodes, E.A.; Robinson, W.R.; Stanford, G.S.; Wright, A.E.

1986-01-01

TREAT tests M2, M3, and M4 were performed to obtain information on two key behavior characteristics of fuel under transient overpower accident conditions in metal-fueled fast reactors: the prefailure axial self-extrusion (elongation beyond thermal expansion) of fuel within intact cladding and the margin to cladding breach. Uranium-5 wt% fissium Experimental Breeder Reactor-II driver fuel pins were used for the tests since they were available as suitable stand-ins for the uranium-plutonium-zirconium ternary fuel, which is the reference fuel of the integral fast reactor (IFR) concept. The ternary fuel will be used in subsequent TREAT tests. Preliminary results from tests M2 and M3 were presented earlier. The present report includes significant advances in analysis as well as additional data from test M4. Test results and analysis have led to the development and validation of pin cladding failure and fuel extrusion models for metallic fuel, within reasonable uncertainties for the uranium-fissium alloy. Concepts involved are straightforward and readily extendable to ternary alloys and behavior in full-size reactors

DESIGN STUDY FOR A LOW-ENRICHED URANIUM CORE FOR THE HIGH FLUX ISOTOPE REACTOR, ANNUAL REPORT FOR FY 2010

Energy Technology Data Exchange (ETDEWEB)

Cook, David Howard [ORNL; Freels, James D [ORNL; Ilas, Germina [ORNL; Jolly, Brian C [ORNL; Miller, James Henry [ORNL; Primm, Trent [ORNL; Renfro, David G [ORNL; Sease, John D [ORNL; Pinkston, Daniel [ORNL

2011-02-01

This report documents progress made during FY 2010 in studies of converting the High Flux Isotope Reactor (HFIR) from high enriched uranium (HEU) fuel to low enriched uranium (LEU) fuel. Conversion from HEU to LEU will require a change in fuel form from uranium oxide to a uranium-molybdenum alloy. With axial and radial grading of the fuel foil and an increase in reactor power to 100 MW, calculations indicate that the HFIR can be operated with LEU fuel with no degradation in performance to users from the current level. Studies are reported of support to a thermal hydraulic test loop design, the implementation of finite element, thermal hydraulic analysis capability, and infrastructure tasks at HFIR to upgrade the facility for operation at 100 MW. A discussion of difficulties with preparing a fuel specification for the uranium-molybdenum alloy is provided. Continuing development in the definition of the fuel fabrication process is described.

METMET fuel with Zirconium matrix alloys

International Nuclear Information System (INIS)

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

2008-01-01

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

Preliminary investigations on the use of uranium silicide targets for fission Mo-99 production

Energy Technology Data Exchange (ETDEWEB)

Cols, H.; Cristini, P.; Marques, R.

1997-08-01

The National Atomic Energy Commission (CNEA) of Argentine Republic owns and operates an installation for production of molybdenum-99 from fission products since 1985, and, since 1991, covers the whole national demand of this nuclide, carrying out a program of weekly productions, achieving an average activity of 13 terabecquerel per week. At present they are finishing an enlargement of the production plant that will allow an increase in the volume of production to about one hundred of terabecquerel. Irradiation targets are uranium/aluminium alloy with 90% enriched uranium with aluminium cladding. In view of international trends held at present for replacing high enrichment uranium (HEU) for enrichment values lower than 20 % (LEU), since 1990 the authors are in contact with the RERTR program, beginning with tests to adapt their separation process to new irradiation target conditions. Uranium silicide (U{sub 3}Si{sub 2}) was chosen as the testing material, because it has an uranium mass per volume unit, so that it allows to reduce enrichment to a value of 20%. CNEA has the technology for manufacturing miniplates of uranium silicide for their purposes. In this way, equivalent amounts of Molybdenum-99 could be obtained with no substantial changes in target parameters and irradiation conditions established for the current process with Al/U alloy. This paper shows results achieved on the use of this new target.

Chapter 1. General information about uranium. 1.5. Mechanical properties

International Nuclear Information System (INIS)

Khakimov, N.; Nazarov, Kh.M.; Mirsaidov, I.U.

2012-01-01

Full text: The mechanical properties of uranium depend on its purity, and which mechanical and thermal regimes are used for reprocessing. The average elasticity module value for cast uranium is 20.5•10"-"2 mega newton/m"2 (20.9•10"-"3 kilogram-force/mm"2), strength limit during tension at room temperature is 372–470 mega newton/m"2 (38–48 kilogram-force/mm"2), strength is increased after hardening from β - and γ - phases; and average rigidity by Brinell 19.6–21.6•10"2 mega newton/m"2 (200–220 kilogram-force/mm"2). Exposure by neutron flux (which occurs in nuclear reactors) changes the physical and mechanical properties of uranium: creeping develops and brittleness increases, goods deformation is observed, which forces the operator to use uranium in nuclear reactors as different uranium alloys.

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)

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

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

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

Distribution of uranium in marine sediments

International Nuclear Information System (INIS)

Ordonez R, E.; Ramirez T, J.J.; Lopez M, J.; Aspiazu, J.; Ruiz F, A.C.; Valero C, N.

2008-01-01

The marine sediments obtained by means of a sampling nucleus in the Gulf of Tehuantepec, Mexico, they have been object of crystallographic and morphological characterization. The PIXE analysis of some samples in study is shown. The normal methodology to carry out the alpha spectroscopy indicates that the sample should be dissolved, but due to the nature of the marine sediments, it thinks about the necessity to make a fractional separation of the sample components. In each stratum of the profile it separates the organic part and the mineral to recover the uranium. It was observed that in the organic phase, the uranium is in two oxidation states (IV and Vl), being necessary the radiochemical separation with a liquid/liquid column chromatographic that uses the di-2-ethyl hexyl phosphoric acid as stationary phase. The uranium compounds extracts are electrodeposited in fine layers on stainless steel disks to carry out the analysis by alpha spectroscopy. The spectroscopic analysis of the uranium indicates us that for each stratum one has a difference marked in the quotient of activities of 234 U/ 238 U that depends on the nature of the studied fraction. These results give us a clear idea about how it is presented the effect of the uranium migration and other radioelements in the biosphere, with what we can determine which are the conditions in that these have their maximum mobility and to know their diffusion patterns in the different media studied. (Author)

Standard test method for analysis of isotopic composition of uranium in nuclear-grade fuel material by quadrupole inductively coupled plasma-mass spectrometry

CERN Document Server

American Society for Testing and Materials. Philadelphia

2000-01-01

1.1 This test method is applicable to the determination of the isotopic composition of uranium (U) in nuclear-grade fuel material. The following isotopic weight percentages are determined using a quadrupole inductively coupled plasma-mass spectrometer (Q-ICP-MS): 233U, 234U, 235U, 236U, and 238U. The analysis can be performed on various material matrices after acid dissolution and sample dilution into water or dilute nitric (HNO3) acid. These materials include: fuel product, uranium oxide, uranium oxide alloys, uranyl nitrate (UNH) crystals, and solutions. The sample preparation discussed in this test method focuses on fuel product material but may be used for uranium oxide or a uranium oxide alloy. Other preparation techniques may be used and some references are given. Purification of the uranium by anion-exchange extraction is not required for this test method, as it is required by other test methods such as radiochemistry and thermal ionization mass spectroscopy (TIMS). This test method is also described i...

Microstructure and mechanical properties of Cu-Ni-Si alloys

International Nuclear Information System (INIS)

Monzen, Ryoichi; Watanabe, Chihiro

2008-01-01

The microstructure and mechanical properties of 0.1 wt.% Mg-added and Mg-free Cu-2.0 wt.% Ni-0.5 wt.% Si alloys aged at 400 deg. C have been examined. The addition of Mg promotes the formation of disk-shaped Ni 2 Si precipitates. The Cu-Ni-Si-Mg alloy exhibits higher strength and resistance to stress relaxation than the Cu-Ni-Si alloy. The higher strength or stress relaxation resistance is attributable to the reduction in inter-precipitate spacing by the Mg addition or the drag effect of Mg atoms on dislocation motion. The Cu-Ni-Si alloy with a large grain size of 150 μm shows higher stress relaxation resistance than the alloy with a small grain size of 10 μm because of a lower density of mobile dislocations in the former alloy

Microstructure and mechanical properties of Cu-Ni-Si alloys

Energy Technology Data Exchange (ETDEWEB)

Monzen, Ryoichi [Division of Innovative Technology and Science, Graduate School of Natural Science and Technology, Kanzawa University, Kakuma-machi, Kanazawa 920-1192 (Japan)], E-mail: monzen@t.kanazawa-u.ac.jp; Watanabe, Chihiro [Division of Innovative Technology and Science, Graduate School of Natural Science and Technology, Kanzawa University, Kakuma-machi, Kanazawa 920-1192 (Japan)

2008-06-15

The microstructure and mechanical properties of 0.1 wt.% Mg-added and Mg-free Cu-2.0 wt.% Ni-0.5 wt.% Si alloys aged at 400 deg. C have been examined. The addition of Mg promotes the formation of disk-shaped Ni{sub 2}Si precipitates. The Cu-Ni-Si-Mg alloy exhibits higher strength and resistance to stress relaxation than the Cu-Ni-Si alloy. The higher strength or stress relaxation resistance is attributable to the reduction in inter-precipitate spacing by the Mg addition or the drag effect of Mg atoms on dislocation motion. The Cu-Ni-Si alloy with a large grain size of 150 {mu}m shows higher stress relaxation resistance than the alloy with a small grain size of 10 {mu}m because of a lower density of mobile dislocations in the former alloy.

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

Development of very high-density low-enriched uranium fuels

International Nuclear Information System (INIS)

Snelgrove, J.L.; Hofman, G.L.; Trybus, C.L.; Wiencek, T.C.

1997-02-01

The RERTR program has recently begun an aggressive effort to develop dispersion fuels for research and test reactors with uranium densities of 8 to 9 g U/cm 3 , based on the use of γ-stabilized uranium alloys. Fabrication development teams and facilities are being put into place and preparations for the first irradiation test are in progress. The first screening irradiations are expected to begin in late April 1997 and first results should be available by end of 1997. Discussions with potential international partners in fabrication development and irradiation testing have begun

Set up of Uranium-Molybdenum powder production (HMD process)

International Nuclear Information System (INIS)

Lopez, Marisol; Pasqualini, Enrique E.; Gonzalez, Alfredo G.

2003-01-01

Powder metallurgy offers different alternatives for the production of Uranium-Molybdenum (UMo) alloy powder in sizes smaller than 150 microns. This powder is intended to be used as a dispersion fuel in an aluminum matrix for research, testing and radioisotopes production reactors (MTR). A particular process of massive hydriding the UMo alloy in gamma phase has been developed. This work describes the final adjustments of process variables to obtain UMo powder by hydriding-milling-de hydriding (HMD) and its capability for industrial scaling up. (author)

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

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

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

International Nuclear Information System (INIS)

Balaramamoorthy, K.

1992-01-01

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

Degradation of bioabsorbable Mg-based alloys: Assessment of the effects of insoluble corrosion products and joint effects of alloying components on mammalian cells

International Nuclear Information System (INIS)

Grillo, Claudia A.; Alvarez, Florencia; Fernández Lorenzo de Mele, Mónica A.

2016-01-01

This work is focused on the processes occurring at the bioabsorbable metallic biomaterial/cell interfaces that may lead to toxicity. A critical analysis of the results obtained when degradable metal disks (pure Mg and rare earth-containing alloys (ZEK100 alloys)) are in direct contact with cell culture and those obtained with indirect methods such as the use of metal salts and extracts was made. Viability was assessed by Acridine Orange dye, neutral red and clonogenic assays. The effects of concentration of corrosion products and possible joint effects of the binary and ternary combinations of La, Zn and Mg ions, as constituents of ZEK alloys, were evaluated on a mammalian cell culture. In all cases more detrimental effects were found for pure Mg than for the alloys. Experiments with disks showed that gradual alterations in pH and in the amount of corrosion products were better tolerated by cells and resulted in higher viability than abrupt changes. In addition, viability was dependent on the distance from the source of ions. Experiments with extracts showed that the effect of insoluble degradation products was highly detrimental. Indirect tests with Zn ions revealed that harmful effects may be found at concentrations ≥ 150 μM and at ≥ 100 μM in mixtures with Mg. These mixtures lead to more deleterious effects than single ions. Results highlight the need to develop a battery of tests to evaluate the biocompatibility of bioabsorbable biomaterials. - Highlights: • A metal disk setup is better in simulating in vivo situations than extracts and salts. • The biodegradation process and cell metabolism were interdependent. • Zn (100 μM) and Mg (8.2 × 10"3 μM) mixtures are more toxic than single Zn or Mg. • Insoluble degradation products of Mg showed high negative effect on cell viability.

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%

Characterization of plasma coated tungsten heavy alloy

International Nuclear Information System (INIS)

Bose, A.; Kapoor, D.; Lankford, J. Jr.; Nicholls, A.E.

1996-01-01

The detrimental environmental impact of Depleted Uranium-based penetrators have led to tremendous development efforts in the area of tungsten heavy alloy based penetrators. One line of investigation involves the coating of tungsten heavy alloys with materials that are prone to shear localization. Plasma spraying of Inconel 718 and 4340 steel have been used to deposit dense coatings on tungsten heavy alloy substrates. The aim of the investigation was to characterize the coating primarily in terms of its microstructure and a special push-out test. The paper describes the results of the push-out tests and analyzes some of the possible failure mechanisms by carrying out microstructural characterization of the failed rings obtained from the push out tests

Evaporation monitoring and composition control of alloy systems with widely differing vapor pressures

International Nuclear Information System (INIS)

Anklam, T.M.; Berzins, L.V.; Braun, D.G.; Haynam, C.; McClelland, M.A.; Meier, T.

1994-10-01

Lawrence Livermore National Laboratory is developing sensors and controls to improve and extend electron beam materials processing technology to alloy systems with constituents of widely varying vapor pressure. The approach under development involves using tunable lasers to measure the density and composition of the vapor plume. A laser based vaporizer control system for vaporization of a uranium-iron alloy has been previously demonstrated in multi-hundred hour, high rate vaporization experiments at LLNL. This paper reviews the design and performance of the uranium vaporization sensor and control system and discusses the extension of the technology to monitoring of uranium vaporization. Data is presented from an experiment in which titanium wire was fed into a molten niobium pool. Laser data is compared to deposited film composition and film cross sections. Finally, the potential for using this technique for composition control in melting applications is discussed

Innovative opto-mechanical design of a laser head for compact thin-disk

Science.gov (United States)

Macúchová, Karolina; Smrž, Martin; Řeháková, Martina; Mocek, Tomáš

2016-11-01

We present recent progress in design of innovative versatile laser head for lasers based on thin-disk architecture which are being constructed at the HiLASE centre of the IOP in the Czech Republic. Concept of thin-disk laser technology allows construction of lasers providing excellent beam quality with high average output power and optical efficiency. Our newly designed thin-disk carrier and pump module comes from optical scheme consisting of a parabolic mirror and roof mirrors proposed in 90's. However, mechanical parts and a cooling system were in-house simplified and tailor-made to medium power lasers since no suitable setup was commercially available. Proposed opto-mechanical design is based on stable yet easily adjustable mechanics. The only water nozzle-cooled component is a room-temperature-operated thindisk mounted on a special cooling finger. Cooling of pump optics was replaced by heat conductive transfer from mirrors made of special Al alloy to a massive brass baseplate. Such mirrors are easy to manufacture and very cheap. Presented laser head was manufactured and tested in construction of Er and Yb doped disk lasers. Details of the latest design will be presented.

Development of very-high-density low-enriched uranium fuels

International Nuclear Information System (INIS)

Snegrove, J.L.; Hofmann, G.L.; Trybus, C.L.; Wiencek, T.C.

1997-01-01

The RERTR (=Reduced Enrichment for Research and Test Reactors) program has begun an aggressive effort to develop dispersion fuels for research and test reactors with uranium densities of 8 to 9 g U/cm 3 , based on the use of γ-stabilized uranium alloys. Fabrication development teams and facilities are being put into place, and preparations for the first irradiation test are in progress. The first screening irradiations are expected to begin in late April 1997 and the first results should be available by the end of 1997. Discussions with potential international partners in fabrication development and irradiation testing have begun. (author)

Magnesium and uranium ignition in different gaseous atmospheres

International Nuclear Information System (INIS)

Darras, R.; Baque, P.; Leclercq, D.

1960-01-01

Magnesium, uranium and some of their alloys burning temperatures have been systematically determined in an air or carbon dioxide atmosphere, either dry or wet. Two different ways of heating have been used: either continuously rising up the temperature, or heating to and then maintaining a constant temperature. The results are clearly different in the two cases. Besides, if moisture has little effect on the magnesium burning temperatures in air, it does lower them by about 130-140 deg. C in CO 2 . The differences of sight between the burning of magnesium and uranium have been noticed; this leads to distinguish between an 'ignition' and an 'inflammation'. (author) [fr

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

Separation and mass spectrometry of nanogram quantities of uranium and thorium from thorium-uranium dioxide fuels

International Nuclear Information System (INIS)

Green, L.W.; Elliot, N.L.; Longhurst, T.H.

1983-01-01

A microchemical procedure was developed for the separation and isotopic analysis of U and Th from irradiated (Th,U)O 2 fuel. The separation procedure consisted of two stages; in the first a tributyl phosphate impregnated resin bead was equilibrated with the dissolved fuel in 0.08 M HF/6 M HNO 3 solution. The bead sorbed approximately 1.7 μg of U and 4.8μg of Th and provided good separation of these from the fission products. In the second stage, the U and Th were back-extracted into 0.025 M HF/8 M HNO 3 solution, which contained a small anion-exchange membrane disk. The disk adsorbed approximately 14 ng of U and 45 ng of Th, and subsequently was transferred to the ionizing filament of a thermal-ionization mass spectrometer and covered with a starch deposit. Sensitivities were sufficiently high for sequential analysis of these quantities of Th and U from a single disk. Isotopic data obtained for a combined U and Th standard showed excellent agreement with certified values: overall bias and precision were < 0.03% and 0.2% relative standard deviation, respectively, for both elements. The applicability of the procedure to uranium fuels was also demonstrated. 6 figures, 2 tables

Separation and mass spectrometry of nanogram quantities of uranium and thorium from thorium-uranium dioxide fuels

Energy Technology Data Exchange (ETDEWEB)

Green, L.W.; Elliot, N.L.; Longhurst, T.H

1983-07-01

A convenient and sensitive microchemical procedure was developed for the separation and isotopic analysis of U and Th from irradiated (Th,U)O{sub 2} fuel. The separation procedure consisted of two stages; in the first a tributyl phosphate impregnated resin bead was equilibrated with the dissolved fuel in 0.08 M HF/6 M HNO{sub 3} solution. The bead sorbed approximately 1.7 {mu}g of U and 4.8 {mu}g of Th and provided good separation of these from the fission products. In the second stage, the U and Th were back-extracted into 0.025 M HF/8 M HNO{sub 3} solution, which contained a small anion-exchange membrane disk. The disk adsorbed approximately 14 ng of U and 45 ng of Th, and subsequently was transferred to the ionizing filament of a thermal-ionization mass spectrometer and covered with a starch deposit. Sensitivities were sufficiently high for sequential analysis of these quantities of Th and U from a single disk. Isotopic data obtained for a combined U and Th standard showed excellent agreement with certified values: overall bias and precision were < -0.03% and 0.2% relative standard deviation, respectively, for both elements. The applicability of the procedure to uranium fuels was also demonstrated. (author)

Fast vortex oscillations in a ferrimagnetic disk near the angular momentum compensation point

Science.gov (United States)

Kim, Se Kwon; Tserkovnyak, Yaroslav

2017-07-01

We theoretically study the oscillatory dynamics of a vortex core in a ferrimagnetic disk near its angular momentum compensation point, where the spin density vanishes but the magnetization is finite. Due to the finite magnetostatic energy, a ferrimagnetic disk of suitable geometry can support a vortex as a ground state similar to a ferromagnetic disk. In the vicinity of the angular momentum compensation point, the dynamics of the vortex resemble those of an antiferromagnetic vortex, which is described by equations of motion analogous to Newton's second law for the motion of particles. Owing to the antiferromagnetic nature of the dynamics, the vortex oscillation frequency can be an order of magnitude larger than the frequency of a ferromagnetic vortex, amounting to tens of GHz in common transition-metal based alloys. We show that the frequency can be controlled either by applying an external field or by changing the temperature. In particular, the latter property allows us to detect the angular momentum compensation temperature, at which the lowest eigenfrequency attains its maximum, by performing ferromagnetic resonance measurements on the vortex disk. Our work proposes a ferrimagnetic vortex disk as a tunable source of fast magnetic oscillations and a useful platform to study the properties of ferrimagnets.

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

Development of advanced P/M Ni-base superalloys for turbine disks

Directory of Open Access Journals (Sweden)

Garibov Genrikh S.

2014-01-01

Full Text Available In the process of evolution of powder metallurgy in Russia the task permanently formulated was the following: to improve strength properties of P/M superalloys without application of additional complex HIPed blanks deformation operation. On the other hand development of a turbine disk material structure to ensure an improvement in aircraft engine performance requires the use of special HIP and heat treatment conditions. To ensure maximum strength properties of disk materials it is necessary to form a structure which would have optimum size of solid solution grains, γ′-phases and carbides. Along with that heating of the material up to a temperature determined by solvus of an alloy ensures a stable and reproducible level of mechanical properties of the disks. The above-said can be illustrated by successful mastering of new complex-alloyed VVP-class superalloys with the use of powder size − 100 μm. Application of special HIP and heat treatment conditions for these superalloys to obtain the desired grain size and the strengthening γ′-phase precipitates allowed a noticeable improvement in ultimate tensile strength and yield strength up to ≥1600 MPa and ≥1200 MPa respectively. 100 hrs rupture strength at 650 ∘C and 750 ∘C was improved up to 1140 MPa and 750 MPa respectively. P/M VVP nickel-base superalloys offer higher characteristics in comparison with many superalloys designed for the same purposes. HIPed disc compacts manufactured from PREP-powder have a homogeneous micro- and macrostructure, a stable level of mechanical properties.

Chitosan patterning on titanium alloys

OpenAIRE

Gilabert Chirivella, Eduardo; Pérez Feito, Ricardo; Ribeiro, Clarisse; Ribeiro, Sylvie; Correia, Daniela; González Martin, María Luisa; Manero Planella, José María; Lanceros Méndez, Senentxu; Gallego Ferrer, Gloria; Gómez Ribelles, José Luis

2017-01-01

Titanium and its alloys are widely used in medical implants because of their excellent properties. However, bacterial infection is a frequent cause of titanium-based implant failure and also compromises its osseointegration. In this study, we report a new simple method of providing titanium surfaces with antibacterial properties by alternating antibacterial chitosan domains with titanium domains in the micrometric scale. Surface microgrooves were etched on pure titanium disks at i...

Studies on the interference of hydrofluoric acid and phosphoric acid in the determination of uranium using Ti(III) reduction method-biamperometry end point

International Nuclear Information System (INIS)

Shiny, T.S.; Rajalakshmi, A.; Phal, D.G.; Charyulu, M.M.; Ramakumar, K.L.

2007-01-01

Accurate and precise determination of uranium in nuclear materials is necessary for chemical quality control as well as for nuclear material accounting purposes. Different types of uranium samples are received for the measurements. Depending upon the nature of the sample dissolution procedure is selected. Mixed oxide samples of uranium and plutonium, for example, are dissolved in nitric acid containing hydrofluoric acid under IR lamp. The fluoride ions are removed by repeated evaporation of the solution. However, some fluoride ions are left in the solutions depending on the conditions of evaporation. Uranium samples and alloy samples are dissolved in dilute hydrochloric acid. The rate of dissolution depends on concentration of acid. Sometimes a mixture of hydrochloric acid and hydrofluoric acid is used for the dissolution metal alloy samples, which may contain silica. Another method of dissolution of these samples is using a mixture of phosphoric acid and 1% hydrofluoric acid. It is necessary to study the interference of hydrofluoric acid and phosphoric acid on the determination of uranium

Alloy Development, Processing and Characterization of Devitrified Titanium Base Microcrystalline Alloys.

Science.gov (United States)

1986-01-01

1.5m wide by injecting the molten alloy onto a rotating copper ’. disk through the orifice at the bottom of the copper crucible under inert gas...icrocrystalline forms [10, 271. 7his technique adopts the combination of a water-cooled cold copper crucible with an arc heating scheme that uses non-consumable...are malted in the cold copper crucible and spun in an inert gas atmosphere. he ribbon produced has a uniform thickness of 20 to SOgm. 5’ -7 -. -F -i

Determination of uranium isotopes in environmental samples by anion exchange in sulfuric and hydrochloric acid media

International Nuclear Information System (INIS)

Popov, L.

2016-01-01

Method for determination of uranium isotopes in various environmental samples is presented. The major advantages of the method are the low cost of the analysis, high radiochemical yields and good decontamination factors from the matrix elements, natural and man-made radionuclides. The separation and purification of uranium is attained by adsorption with strong base anion exchange resin in sulfuric and hydrochloric acid media. Uranium is electrodeposited on a stainless steel disk and measured by alpha spectrometry. The analytical method has been applied for the determination of concentrations of uranium isotopes in mineral, spring and tap waters from Bulgaria. The analytical quality was checked by analyzing reference materials. - Highlights: • The method allows cost-effective determination of U isotopes. • High amounts of environmental samples can be analyzed. • High chemical yields, energy resolution and decontamination factors were achieved. • Uranium isotope concentrations in mineral waters from Bulgaria are presented.

The corrosion of depleted uranium in terrestrial and marine environments

International Nuclear Information System (INIS)

Toque, C.; Milodowski, A.E.; Baker, A.C.

2014-01-01

Depleted Uranium alloyed with titanium is used in armour penetrating munitions that have been fired in a number of conflict zones and testing ranges including the UK ranges at Kirkcudbright and Eskmeals. The study presented here evaluates the corrosion of DU alloy cylinders in soil on these two UK ranges and in the adjacent marine environment of the Solway Firth. The estimated mean initial corrosion rates and times for complete corrosion range from 0.13 to 1.9 g cm −2 y −1 and 2.5–48 years respectively depending on the particular physical and geochemical environment. The marine environment at the experimental site was very turbulent. This may have caused the scouring of corrosion products and given rise to a different geochemical environment from that which could be easily duplicated in laboratory experiments. The rate of mass loss was found to vary through time in one soil environment and this is hypothesised to be due to pitting increasing the surface area, followed by a build up of corrosion products inhibiting further corrosion. This indicates that early time measurements of mass loss or corrosion rate may be poor indicators of late time corrosion behaviour, potentially giving rise to incorrect estimates of time to complete corrosion. The DU alloy placed in apparently the same geochemical environment, for the same period of time, can experience very different amounts of corrosion and mass loss, indicating that even small variations in the corrosion environment can have a significant effect. These effects are more significant than other experimental errors and variations in initial surface area. -- Highlights: ► In-situ experiments were conducted to evaluate corrosion rates of depleted uranium. ► Samples were corroded in marine sediments, open sea water and two terrestrial soils. ► The depleted uranium titanium alloy corroded fastest in the marine environments. ► Rates of mass loss can vary through time if corrosion products are not removed. �

Identifying Likely Disk-hosting M dwarfs with Disk Detective

Science.gov (United States)

Silverberg, Steven; Wisniewski, John; Kuchner, Marc J.; Disk Detective Collaboration

2018-01-01

M dwarfs are critical targets for exoplanet searches. Debris disks often provide key information as to the formation and evolution of planetary systems around higher-mass stars, alongside the planet themselves. However, less than 300 M dwarf debris disks are known, despite M dwarfs making up 70% of the local neighborhood. The Disk Detective citizen science project has identified over 6000 new potential disk host stars from the AllWISE catalog over the past three years. Here, we present preliminary results of our search for new disk-hosting M dwarfs in the survey. Based on near-infrared color cuts and fitting stellar models to photometry, we have identified over 500 potential new M dwarf disk hosts, nearly doubling the known number of such systems. In this talk, we present our methodology, and outline our ongoing work to confirm systems as M dwarf disks.

Endurance in Al Alloy Melts and Wear Resistance of Titanium Matrix Composite Shot-Sleeve for Aluminum Alloy Die-casting

International Nuclear Information System (INIS)

Choi, Bong-Jae; Kim, Young-Jig; Sung, Si-Young

2012-01-01

The main purpose of this study was to evaluate the endurance against Al alloy melts and wear resistance of an in-situ synthesized titanium matrix composite (TMC) sleeve for aluminum alloy die-casting. The conventional die-casting shot sleeve material was STD61 tool steel. TMCs have great thermal stability, wear and oxidation resistance. The in-situ reaction between Ti and B4C leads to two kinds of thermodynamically stable reinforcements, such as TiBw and TiCp. To evaluate the feasibility of the application to a TMCs diecasting shot sleeve, the interfacial reaction behavior was examined between Al alloys melts with TMCs and STD61 tool steel. The pin-on-disk type dry sliding wear test was also investigated for TMCs and STD61 tool steel.

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

ALMA Survey of Lupus Protoplanetary Disks. II. Gas Disk Radii

Science.gov (United States)

Ansdell, M.; Williams, J. P.; Trapman, L.; van Terwisga, S. E.; Facchini, S.; Manara, C. F.; van der Marel, N.; Miotello, A.; Tazzari, M.; Hogerheijde, M.; Guidi, G.; Testi, L.; van Dishoeck, E. F.

2018-05-01

We present Atacama Large Millimeter/Sub-Millimeter Array (ALMA) Band 6 observations of a complete sample of protoplanetary disks in the young (∼1–3 Myr) Lupus star-forming region, covering the 1.33 mm continuum and the 12CO, 13CO, and C18O J = 2–1 lines. The spatial resolution is ∼0.″25 with a medium 3σ continuum sensitivity of 0.30 mJy, corresponding to M dust ∼ 0.2 M ⊕. We apply Keplerian masking to enhance the signal-to-noise ratios of our 12CO zero-moment maps, enabling measurements of gas disk radii for 22 Lupus disks; we find that gas disks are universally larger than millimeter dust disks by a factor of two on average, likely due to a combination of the optically thick gas emission and the growth and inward drift of the dust. Using the gas disk radii, we calculate the dimensionless viscosity parameter, α visc, finding a broad distribution and no correlations with other disk or stellar parameters, suggesting that viscous processes have not yet established quasi-steady states in Lupus disks. By combining our 1.33 mm continuum fluxes with our previous 890 μm continuum observations, we also calculate the millimeter spectral index, α mm, for 70 Lupus disks; we find an anticorrelation between α mm and millimeter flux for low-mass disks (M dust ≲ 5), followed by a flattening as disks approach α mm ≈ 2, which could indicate faster grain growth in higher-mass disks, but may also reflect their larger optically thick components. In sum, this work demonstrates the continuous stream of new insights into disk evolution and planet formation that can be gleaned from unbiased ALMA disk surveys.

Irradiation tests of 99Mo isotope production targets employing uranium metal foils

International Nuclear Information System (INIS)

Hofman, G.L.; Wiencek, T.C.; Wood, E.L.; Snelgrove, J.L.; Suripto, A.; Nasution, H.; Lufti-Amin, D.; Gogo, A.

1996-01-01

Most of the world's supply of 99 mTc for medical purposes is currently produced from the decay of 99 Mo derived from the fissioning of high-enriched uranium (HEU). Substitution of low-enriched uranium (LEU) metal foils for the HEU UO 2 used in current target designs will allow equivalent 99 Mo yields with little change in target geometries. Substitution of uranium metal for uranium alloy and aluminide in other target designs will also allow the conversion of HEU to LEU. Several uranium-metal-foil targets have been fabricated at ANL and irradiated to prototypic burnup in the Indonesian RSG-GAS reactor. Postirradiation examination of the initial test indicated that design modifications were required to allow the irradiated foil to be removed for chemical processing. The latest test has shown good irradiation behavior, satisfactory dismantling and foil removal when the U-foil is separated from its containment by metallic, fission-recoil absorbing barriers. (author)

Irradiation tests of 99Mo isotope production targets employing uranium metal foils

International Nuclear Information System (INIS)

Hofman, G.L.; Wiencek, T.C.; Wood, E.L.; Snelgrove, J.L.; Suripto, A.; Nasution, H.; Lufti-Amin, D.; Gogo, A.

1996-01-01

Most of the world's supply of 99m Tc for medical purposes is currently produced form the decay of 99 Mo derived from the fissioning of high-enriched uranium (HEU). Substitution of low-enriched uranium (LEU) metal foils for the HEU UO 2 used in current target designs will allow equivalent 99 Mo yields with little change in target geometries. Substitution of uranium metal for uranium alloy and aluminide in other target designs will also allow the conversion of HEU to LEU. Several uranium-metal-foil targets have been fabricated at ANL and irradiated to prototypic burnup in the Indonesian RSG-GAS reactor. Postirradiation examination of the initial test indicated that design modifications were required to allow the irradiated foil to be removed for chemical processing. The latest test has shown good irradiation behavior, satisfactory dismantling and foil removal when the U-foil is separated from its containment by metallic, fission-recoil absorbing barriers

THICK-DISK EVOLUTION INDUCED BY THE GROWTH OF AN EMBEDDED THIN DISK

International Nuclear Information System (INIS)

Villalobos, Alvaro; Helmi, Amina; Kazantzidis, Stelios

2010-01-01

We perform collisionless N-body simulations to investigate the evolution of the structural and kinematical properties of simulated thick disks induced by the growth of an embedded thin disk. The thick disks used in the present study originate from cosmologically common 5:1 encounters between initially thin primary disk galaxies and infalling satellites. The growing thin disks are modeled as static gravitational potentials and we explore a variety of growing-disk parameters that are likely to influence the response of thick disks. We find that the final thick-disk properties depend strongly on the total mass and radial scale length of the growing thin disk, and much less sensitively on its growth timescale and vertical scale height as well as the initial sense of thick-disk rotation. Overall, the growth of an embedded thin disk can cause a substantial contraction in both the radial and vertical direction, resulting in a significant decrease in the scale lengths and scale heights of thick disks. Kinematically, a growing thin disk can induce a notable increase in the mean rotation and velocity dispersions of thick-disk stars. We conclude that the reformation of a thin disk via gas accretion may play a significant role in setting the structure and kinematics of thick disks, and thus it is an important ingredient in models of thick-disk formation.

Determination of uranium traces in fuel cans of nuclear reactors; Determinacion de trazas de uranio en vainas de combustible de reactores nucleares

Energy Technology Data Exchange (ETDEWEB)

Acosta L, C.E.; Benavides M, A.M.; Sanchez P, L.A.; Nava S, G.F. [Instituto Nacional de Investigaciones Nucleares, A.P. 18-1027, 11801 Mexico D.F. (Mexico)

1997-07-01

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)

Effect of Microstructure on Time Dependent Fatigue Crack Growth Behavior In a P/M Turbine Disk Alloy

Science.gov (United States)

Telesman, Ignacy J.; Gabb, T. P.; Bonacuse, P.; Gayda, J.

2008-01-01

A study was conducted to determine the processes which govern hold time crack growth behavior in the LSHR disk P/M superalloy. Nineteen different heat treatments of this alloy were evaluated by systematically controlling the cooling rate from the supersolvus solutioning step and applying various single and double step aging treatments. The resulting hold time crack growth rates varied by more than two orders of magnitude. It was shown that the associated stress relaxation behavior for these heat treatments was closely correlated with the crack growth behavior. As stress relaxation increased, the hold time crack growth resistance was also increased. The size of the tertiary gamma' in the general microstructure was found to be the key microstructural variable controlling both the hold time crack growth behavior and stress relaxation. No relationship between the presence of grain boundary M23C6 carbides and hold time crack growth was identified which further brings into question the importance of the grain boundary phases in determining hold time crack growth behavior. The linear elastic fracture mechanics parameter, Kmax, is unable to account for visco-plastic redistribution of the crack tip stress field during hold times and thus is inadequate for correlating time dependent crack growth data. A novel methodology was developed which captures the intrinsic crack driving force and was able to collapse hold time crack growth data onto a single curve.

Oscillations of disks

CERN Document Server

Kato, Shoji

2016-01-01

This book presents the current state of research on disk oscillation theory, focusing on relativistic disks and tidally deformed disks. Since the launch of the Rossi X-ray Timing Explorer (RXTE) in 1996, many high-frequency quasiperiodic oscillations (HFQPOs) have been observed in X-ray binaries. Subsequently, similar quasi-periodic oscillations have been found in such relativistic objects as microquasars, ultra-luminous X-ray sources, and galactic nuclei. One of the most promising explanations of their origin is based on oscillations in relativistic disks, and a new field called discoseismology is currently developing. After reviewing observational aspects, the book presents the basic characteristics of disk oscillations, especially focusing on those in relativistic disks. Relativistic disks are essentially different from Newtonian disks in terms of several basic characteristics of their disk oscillations, including the radial distributions of epicyclic frequencies. In order to understand the basic processes...

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)

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

Mössbauer spectroscopic studies in U-Fe and U-Fe-Zr alloys

Energy Technology Data Exchange (ETDEWEB)

Panda, Alaka; Singh, L. Herojit; Rajagopalan, S.; Govindaraj, R., E-mail: govind@igcar.gov.in; Ramachandran, Renjith; Kalavathi, S.; Amarendra, G. [Materials Science Group, Indira Gandhi Centre for Atomic Research, Kalpakkam (India)

2016-05-23

{sup 57}Fe Mössbauer studies have been carried out in an alloy of U and Fe with atomic percentage in the ratio of 68%:32% in order to understand the local structure and valence of Fe atoms associated with different phases that may get formed. The effect of changes in the hyperfine parameters such as isomer shift and quadrupole splitting at Fe sites due to additional alloying of Zr has been studied in an alloy of U, Fe and Zr in the ratio of 44%:33%:23% respectively with respect to that of the U-Fe alloy chosen in the present study. Possible effect of solute clustering in these systems has been addressed in an analogous alloy of uranium and zirconium using positron lifetime spectroscopy.

Estimation of formation heat of rare earth and actinide alloys

International Nuclear Information System (INIS)

Shubin, A.B.; Yamshchikov, L.F.; Raspopin, S.P.

1986-01-01

A method for forecasting the enthalpy of formation of scandium, yttrium, lanthanum and lanthanides, thorium, uranium and plutonium alloys with a series of fusible metals (Al, Ga, In, Tl, Sn, Pb, Sb, Bi) is proposed. The obtained confidence internal value for the calculated Δ f H 0 values exceeds sufficiently the random error of the experimental determination of the rare metal alloy formation enthalpies. However, taking into account considerable divergences in results of Δ f H 0 determinations performed by different science groups, one may conclude, that such forecasting accuracy may be useful in the course of estimation calculations, especially, for actinide element alloys

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

Magnetically Induced Disk Winds and Transport in the HL Tau Disk

International Nuclear Information System (INIS)

Hasegawa, Yasuhiro; Flock, Mario; Turner, Neal J.; Okuzumi, Satoshi

2017-01-01

The mechanism of angular momentum transport in protoplanetary disks is fundamental to understanding the distributions of gas and dust in the disks. The unprecedented ALMA observations taken toward HL Tau at high spatial resolution and subsequent radiative transfer modeling reveal that a high degree of dust settling is currently achieved in the outer part of the HL Tau disk. Previous observations, however, suggest a high disk accretion rate onto the central star. This configuration is not necessarily intuitive in the framework of the conventional viscous disk model, since efficient accretion generally requires a high level of turbulence, which can suppress dust settling considerably. We develop a simplified, semi-analytical disk model to examine under what condition these two properties can be realized in a single model. Recent, non-ideal MHD simulations are utilized to realistically model the angular momentum transport both radially via MHD turbulence and vertically via magnetically induced disk winds. We find that the HL Tau disk configuration can be reproduced well when disk winds are properly taken into account. While the resulting disk properties are likely consistent with other observational results, such an ideal situation can be established only if the plasma β at the disk midplane is β 0 ≃ 2 × 10 4 under the assumption of steady accretion. Equivalently, the vertical magnetic flux at 100 au is about 0.2 mG. More detailed modeling is needed to fully identify the origin of the disk accretion and quantitatively examine plausible mechanisms behind the observed gap structures in the HL Tau disk.

Magnetically Induced Disk Winds and Transport in the HL Tau Disk

Energy Technology Data Exchange (ETDEWEB)

Hasegawa, Yasuhiro; Flock, Mario; Turner, Neal J. [Jet Propulsion Laboratory, California Institute of Technology, Pasadena, CA 91109 (United States); Okuzumi, Satoshi, E-mail: yasuhiro@caltech.edu [Department of Earth and Planetary Sciences, Tokyo Institute of Technology, Meguro-ku, Tokyo 152-8551 (Japan)

2017-08-10

The mechanism of angular momentum transport in protoplanetary disks is fundamental to understanding the distributions of gas and dust in the disks. The unprecedented ALMA observations taken toward HL Tau at high spatial resolution and subsequent radiative transfer modeling reveal that a high degree of dust settling is currently achieved in the outer part of the HL Tau disk. Previous observations, however, suggest a high disk accretion rate onto the central star. This configuration is not necessarily intuitive in the framework of the conventional viscous disk model, since efficient accretion generally requires a high level of turbulence, which can suppress dust settling considerably. We develop a simplified, semi-analytical disk model to examine under what condition these two properties can be realized in a single model. Recent, non-ideal MHD simulations are utilized to realistically model the angular momentum transport both radially via MHD turbulence and vertically via magnetically induced disk winds. We find that the HL Tau disk configuration can be reproduced well when disk winds are properly taken into account. While the resulting disk properties are likely consistent with other observational results, such an ideal situation can be established only if the plasma β at the disk midplane is β {sub 0} ≃ 2 × 10{sup 4} under the assumption of steady accretion. Equivalently, the vertical magnetic flux at 100 au is about 0.2 mG. More detailed modeling is needed to fully identify the origin of the disk accretion and quantitatively examine plausible mechanisms behind the observed gap structures in the HL Tau disk.

Mechanical and structural behaviour of uranium α, β, γ phases during plastic deformation

International Nuclear Information System (INIS)

Prunier, C.; Collot, C.

1981-06-01

High temperature behaviour of rich and poor uranium alloys in α, β and γ crystalline structures is studied: dynamic recrystallization phenomena begins only in α and β phases high temperature range, high strength and brittle β phase shows a very large ductility above 700 0 C. Dynamic recrystallization in γ phase rich alloys is observed only if large energy is available. Recrystallization is a thermal actived phenomena localised at grain boundary, dependant with alloy concentration and crystalline structure. β phase activation energy and deformation rate for dynamic recrystallization beginning are the most important in relation with structure complexity; both temperature and rate deformation are dynamic recrystallization factors [fr

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

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)

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)

INTURGEO: The international uranium geology information system

International Nuclear Information System (INIS)

1988-09-01

The International Uranium Geology Information System (INTURGEO) is an international compilation of data on uranium deposits and occurrences. The purpose of INTURGEO is to provide a clearinghouse for uranium geological information that can serve for the better understanding of the worldwide distribution of uranium occurrences and deposits. The INTURGEO system is by no means complete for all regions of the world. Data have been available principally from the WOCA countries. INTURGEO currently covers 6,089 occurrences and deposits in 96 countries of which 4,596 occurrences in 92 countries are presented here. The information presented in this publication is a very brief, one line synopsis of deposits and occurrences, and has been collected from literature and through questionnaires sent directly to IAEA Member States. None of the information contained in the INTURGEO database was derived from confidential sources although there are many entries which come from the internal files of Member States and are not directly available in the general literature. The uniformity of the INTURGEO data presented in this report has depended heavily on the data provided by Member States. Basic information includes the deposit or occurrence name, the mining district, the tectonic setting, the geological type, status, size, host-rock type, age of mineralization and bibliographic references. The data contained in the maps of the atlas include all reported occurrences of uranium above the anomaly level. The categories of occurrence and deposit status includes: Anomaly; occurrences of unknown status; occurrences; prospects; developed prospects; subeconomic deposits; economic deposits; mines; inactive mines; depleted mines. A microcomputer version of INTURGEO on 21 Megabyte Bernoulli disks is available. 5 tabs, 102 maps

Degradation of bioabsorbable Mg-based alloys: Assessment of the effects of insoluble corrosion products and joint effects of alloying components on mammalian cells.

Science.gov (United States)

Grillo, Claudia A; Alvarez, Florencia; Fernández Lorenzo de Mele, Mónica A

2016-01-01

This work is focused on the processes occurring at the bioabsorbable metallic biomaterial/cell interfaces that may lead to toxicity. A critical analysis of the results obtained when degradable metal disks (pure Mg and rare earth-containing alloys (ZEK100 alloys)) are in direct contact with cell culture and those obtained with indirect methods such as the use of metal salts and extracts was made. Viability was assessed by Acridine Orange dye, neutral red and clonogenic assays. The effects of concentration of corrosion products and possible joint effects of the binary and ternary combinations of La, Zn and Mg ions, as constituents of ZEK alloys, were evaluated on a mammalian cell culture. In all cases more detrimental effects were found for pure Mg than for the alloys. Experiments with disks showed that gradual alterations in pH and in the amount of corrosion products were better tolerated by cells and resulted in higher viability than abrupt changes. In addition, viability was dependent on the distance from the source of ions. Experiments with extracts showed that the effect of insoluble degradation products was highly detrimental. Indirect tests with Zn ions revealed that harmful effects may be found at concentrations ≥ 150 μM and at ≥ 100 μM in mixtures with Mg. These mixtures lead to more deleterious effects than single ions. Results highlight the need to develop a battery of tests to evaluate the biocompatibility of bioabsorbable biomaterials. Copyright © 2015 Elsevier B.V. All rights reserved.

Production of uranium-molybdenum particles by spark-erosion

International Nuclear Information System (INIS)

Cabanillas, E.D.; Lopez, M.; Pasqualini, E.E.; Cirilo Lombardo, D.J.

2004-01-01

With the spark-erosion method we have produced spheroidal particles of an uranium-molybdenum alloy using pure water as dielectric. The particles were characterized by optical metallography, scanning electron microscopy, energy dispersive spectrometry and X-ray diffraction. Mostly spherical particles of UO 2 with a distinctive size distribution with peaks centered at 70 and 10 μm were obtained. The particles have central inclusions of U and Mo compounds

Production of uranium-molybdenum particles by spark-erosion

Energy Technology Data Exchange (ETDEWEB)

Cabanillas, E.D. E-mail: cabanill@cnea.gov.ar; Lopez, M.; Pasqualini, E.E.; Cirilo Lombardo, D.J

2004-01-01

With the spark-erosion method we have produced spheroidal particles of an uranium-molybdenum alloy using pure water as dielectric. The particles were characterized by optical metallography, scanning electron microscopy, energy dispersive spectrometry and X-ray diffraction. Mostly spherical particles of UO{sub 2} with a distinctive size distribution with peaks centered at 70 and 10 {mu}m were obtained. The particles have central inclusions of U and Mo compounds.

Contact statuses between functionally graded brake disk and pure pad disk

International Nuclear Information System (INIS)

Shahzamanian, M.M.; Sahari, B.B.; Bayat, M.; Mustapha, F.; Ismarrubie, Z.N.; Shahrjerdi, A.

2009-01-01

Full text: The contact statuses between functionally graded (FG) brake disks and pure pad disk are investigated by using finite element method (FEM). Two types of variation is considered for FG brake disk, the variation of materials are considered change in radial and thickness direction of disk. The material properties of these two types of FG brake disks are assumed to be represented by power-law distributions in the radius and thickness direction. The results are obtained and then compared. For the radial FG brake disk, the inner and outer surfaces are considered metal and ceramic respectively, and friction coefficient between metal surface and ceramic surface of FG brake dick with pad are considered 1.4 and 0.75 respectively. For the thickness FG brake disk the contact surface with pure pad brake disk is ceramic and the free surface is metal and friction coefficient between ceramic (contact) surface and pure pad brake disk is considered 0.75. In both types of FG brake disks the Coulomb contact friction is applied. Mechanical response of FG brake disks are compared and verified with the known results in the literatures. Three types of contact statuses are introduced as Sticking, Contact and Near Contact. The contact status between pad and disk for different values for pad thickness, grading index,n , and percentage of friction coefficient (λ) is shown. It can be seen that for all values of percentage of friction coefficient,λ , and grading indices, n, by increasing the thickness of pad cause the contact status changes from sticking to contact and then to near contact. (author)

Development of a technique of preparation of uranium screens for soft X Ray spectrography (1960)

International Nuclear Information System (INIS)

Bersuder, L. de

1961-01-01

The present work concerns the preparation of thin layers of pure uranium (thickness 100 to 1000 Angstrom) by thermal evaporation under vacuum. The protection of uranium against oxidation is obtained by using aluminium deposits under and above the uranium layer. The purity of the layers obtained is checked by electron diffraction and the necessary conditions to avoid oxidation and alloy formation during the formation of deposit are studied. Three methods for measuring the thickness are examined: by α particle counting, by weighing the condensed mass and by weighing the evaporated mass. The method using α particle counting turned to be the most accurate for low thickness layers. (author) [fr

Mechanical behaviour of Nd-Fe-B alloys in the semi-solid state

International Nuclear Information System (INIS)

Oliveira, I.L.; Sinka, V.; Ferrante, M.

1996-01-01

Two alloys with composition Nd 17.6 Fe 75.3 B1.2 Cu 5.9 and Nd 15.9 Fe 77.7 B 5 Cu 1.4 were vacuum induction melted and cast into cylindrical ingots. Samples with 12.3 and 13 mm diameter were deformed with different rates and deformation ratios. One alloy was deformed at 800 deg C between two parallel disks under constant load. Results show that these alloys behave as no-Newtonian fluids. This fact gives a better understanding of both magnetic and crystallographic texture development. Also, changes were detected in the behaviour of semisolid in the course of deformation. (author)

Study of the aqueous chemical treatment of uranium zirconium fuels; Etude du traitement chimique des combustibles uraniumzirconium par voie seche

Energy Technology Data Exchange (ETDEWEB)

Bourgeois, M; Nollet, P [Commissariat a l' Energie Atomique, Fontenay-aux-Roses (France). Centre d' Etudes Nucleaires

1964-07-01

A dry process has been studied for separating the uranium from the zirconium-either for recovering the enriched uranium from fuel element production waste, or with a view to treating this waste after irradiation. In this process the alloy is treated with hydrochloric acid at 400 deg. C in a fluidized corundum bed which causes the zirconium to volatilize as tetrachloride and the uranium to form the trichloride. This latter is then converted to the hexafluoride by attack with fluorure. After the laboratory tests, a first pilot plant with a capacity of 1 kg of alloy was tried out at the Fontenay-aux-Roses Nuclear Research Centre; this made it possible to fix the operational conditions for the process. An industrial scale plant was then built with the collaboration of the from Kuhlmann, and operated until a satisfactory process had been developed for treating the waste. This installation treats 3 kg/h of alloy with a yield for the hydrochloric acid of about 50 per cent and with a uranium loss in the zirconium tetrachloride of about 0.1 per cent. An active pilot plant capable of treating of treating a few kilos of irradiated alloy is now being studied. (authors) [French] On a etudie un procede de voie seche pour effectuer la separation de l'uranium et du zirconium - soit en vue de la recuperation de l'uranium enrichi contenu dans les dechets de fabrication des elements combustibles - soit en vue du traitement de ceux-ci apres irradiation. Ce procede consiste a attaquer l'alliage par l'acide chlorhydrique a 400 deg. C dans un lit fluidise de corindon, ce qui a pour effet de volatiliser le zirconium sous forme de tetrachlorure et de transformer l'uranium en trichlorure. Ce dernier est ensuite converti en hexafluorure par action du fluor. Apres des essais de laboratoire, un premier pilote a l'echelle de 1 kg d'alliage a ete experimente au Centre d'Etudes Nucleaires de Fontenay-aux-Roses et a permis de determiner les conditions operatoires du procede. En collaboration avec

GIANT PLANET MIGRATION, DISK EVOLUTION, AND THE ORIGIN OF TRANSITIONAL DISKS

International Nuclear Information System (INIS)

Alexander, Richard D.; Armitage, Philip J.

2009-01-01

We present models of giant planet migration in evolving protoplanetary disks. Our disks evolve subject to viscous transport of angular momentum and photoevaporation, while planets undergo Type II migration. We use a Monte Carlo approach, running large numbers of models with a range in initial conditions. We find that relatively simple models can reproduce both the observed radial distribution of extrasolar giant planets, and the lifetimes and accretion histories of protoplanetary disks. The use of state-of-the-art photoevaporation models results in a degree of coupling between planet formation and disk clearing, which has not been found previously. Some accretion across planetary orbits is necessary if planets are to survive at radii ∼<1.5 AU, and if planets of Jupiter mass or greater are to survive in our models they must be able to form at late times, when the disk surface density in the formation region is low. Our model forms two different types of 'transitional' disks, embedded planets and clearing disks, which show markedly different properties. We find that the observable properties of these systems are broadly consistent with current observations, and highlight useful observational diagnostics. We predict that young transition disks are more likely to contain embedded giant planets, while older transition disks are more likely to be undergoing disk clearing.

Tensile strength of two soldered alloys (Minalux and Verabond2

Directory of Open Access Journals (Sweden)

Mir Mohammad Rezaee S

2002-07-01

Full Text Available Recently. Minalux alloy, a base metal free from Be, has been presented on the market while no special soldering has been recommended for it. On the other hand, based on the manufacturer's claim, this alloy is similar to Verabond2. The aim of this study was to investigate the tensile strength of Minalux and Verabond2, soldered by Verasolder. Twelve standard dambble shape samples, with the length of 18 mm and the diameter of 3mm, were prepared from each alloy. Six samples of each alloy were divided into two pieces with carboradom disk. Soldering gap distance was 0.3mm, measured by a special jig and they were soldered by Verasolder alloy. Six other samples, of both Iranian and foreign unsoldered alloys were considered as control group. Then samples were examined under tensile force and their tensile strength was recorded. Two- way variance analysis showed that the tensile strength of Minalux alloy and Verabond2 were not statistically significant (Verasoler 686, Minalux 723, but after soldering, such difference became significant (Minalux 308, Verabond2 432. Verabond2 showed higher tensile strength after soldering.

Contribution to the study of the hydrolysis of uranium carbides (1963); Contribution a l'etude de l'hydrolyse des carbures d'uranium (1963)

Energy Technology Data Exchange (ETDEWEB)

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

1963-06-15

The hydrolysis of uranium monocarbide in neutral or acid medium leads to the formation of a complex mixture of hydrogen and hydrocarbons mostly saturated. When UC-U alloys are dissolved in hydrochloric-phosphoric medium, the free uranium contents can be determined with good accuracy from the composition of the gaseous phase. The hydrolysis of mixtures of uranium mono - and dicarbide in neutral or acid medium, leads to the formation of a complex mixture of hydrogen and gaseous and condensed hydrocarbons, the composition of which is principally dependent upon the UC{sub 2} content. The reaction mechanisms which are presented in this paper for the hydrolysis of UC and UC{sub 2} provide account for all experimental observations. (author) [French] L'hydrolyse en milieu neutre ou acide du monocarbure d'uranium conduit a la formation d'un melange complexe d'hydrogene et d'hydrocarbures, satures en grande majorite. L'attaque en milieu chlorhydrique-phosphorique des alliages UC-U permet la determination avec une bonne precision, des teneurs en uranium libre a partir de la composition des gaz degages. L'hydrolyse en milieu neutre ou acide des melanges de mono - et dicarbure d'uranium conduit a la formation d'un melange complexe d'hydrogene et d'hydrocarbures gazeux et condenses, dont la composition est essentiellement fonction de la teneur en UC{sub 2}. Les mecanismes reactionnels proposes pour l'hydrolyse de UC et UC{sub 2} rendent compte de tous les faits experimentaux observes. (auteur)

CT-guided percutaneous laser disk decompression for cervical and lumbar disk hernia

International Nuclear Information System (INIS)

Shimizu, Kanichiro; Koyama, Tutomu; Harada, Junta; Abe, Toshiaki

2008-01-01

Percutaneous laser disk decompression under X-ray fluoroscopy was first reported in 1987 for minimally invasive therapy of lumbar disk hernia. In patients with disk hernia, laser vaporizes a small portion of the intervertebral disk thereby reducing the volume and pressure of the affected disk. We present the efficacy and safety of this procedure, and analysis of fair or poor response cases. In our study, 226 cases of lumbar disk hernia and 7 cases of cervical disk hernia were treated under CT guided PLDD. Japan Orthopedic Association (JOA) score and Mac-Nab criteria were investigated to evaluate the response to treatment. Improvement ratio based on the JOA score was calculated as follows. Overall success rate was 91.6% in cases lumber disk hernia, and 100% in cases of cervical disk hernia. We experienced two cases with two cases with postoperative complication. Both cases were treated conservatively. The majority of acute cases and post operative cases were reported to be 'good' on Mac-Nab criteria. Cases of fair or poor response on Mac-Nab criteria were lateral type, foraminal stenosis or large disk hernia. CT-guided PLDD is a safe and accurate procedure. The overall success rate can be increased by carefully selecting patients. (author)

Disk Storage Server

CERN Multimedia

This model was a disk storage server used in the Data Centre up until 2012. Each tray contains a hard disk drive (see the 5TB hard disk drive on the main disk display section - this actually fits into one of the trays). There are 16 trays in all per server. There are hundreds of these servers mounted on racks in the Data Centre, as can be seen.

Some characteristics of the air in a uranium mine

International Nuclear Information System (INIS)

Renoux, A.; Barzic, J.Y.; Madelaine, G.J.; Zettwoog, P.

1978-01-01

The radon content in the atmosphere of a uranium mine, 183 pCi l -1 , was found during the varied phases of the excavation (drilling, blasting, and clearing) to vary between 63 and 3600 pCi l -1 . Radioactive equilibrium was not found to be reached for radon and its daughter products. By means of a seven-stage Andersen cascade impactor, the particle size distribution for the aerosols of the mine was determined as well as the alpha-particle activities on each disk of the impactor and on the millipore filter placed behind each stage. This yielded the information that the major portion of alpha activity in the test mine is connected with aerosols having a radius 1 μm) is very small (<3%). This indicates that if the Andersen impactor is used carelessly, it may yield an erroneous distribution of the radioactivity in a uranium mine. 11 tables. 13 figures

HNC IN PROTOPLANETARY DISKS

International Nuclear Information System (INIS)

Graninger, Dawn; Öberg, Karin I.; Qi, Chunhua; Kastner, Joel

2015-01-01

The distributions and abundances of small organics in protoplanetary disks are potentially powerful probes of disk physics and chemistry. HNC is a common probe of dense interstellar regions and the target of this study. We use the Submillimeter Array (SMA) to observe HNC 3–2 toward the protoplanetary disks around the T Tauri star TW Hya and the Herbig Ae star HD 163296. HNC is detected toward both disks, constituting the first spatially resolved observations of HNC in disks. We also present SMA observations of HCN 3–2 and IRAM 30 m observations of HCN and HNC 1–0 toward HD 163296. The disk-averaged HNC/HCN emission ratio is 0.1–0.2 toward both disks. Toward TW Hya, the HNC emission is confined to a ring. The varying HNC abundance in the TW Hya disk demonstrates that HNC chemistry is strongly linked to the disk physical structure. In particular, the inner rim of the HNC ring can be explained by efficient destruction of HNC at elevated temperatures, similar to what is observed in the ISM. However, to realize the full potential of HNC as a disk tracer requires a combination of high SNR spatially resolved observations of HNC and HCN and disk-specific HNC chemical modeling

HNC IN PROTOPLANETARY DISKS

Energy Technology Data Exchange (ETDEWEB)

Graninger, Dawn; Öberg, Karin I.; Qi, Chunhua [Harvard-Smithsonian Center for Astrophysics, 60 Garden Street, Cambridge, MA 02138 (United States); Kastner, Joel, E-mail: dgraninger@cfa.harvard.edu [Center for Imaging Science, School of Physics and Astronomy, and Laboratory for Multiwavelength Astrophysics, Rochester Institute of Technology, 54 Lomb Memorial Drive, Rochester, NY 14623 (United States)

2015-07-01

The distributions and abundances of small organics in protoplanetary disks are potentially powerful probes of disk physics and chemistry. HNC is a common probe of dense interstellar regions and the target of this study. We use the Submillimeter Array (SMA) to observe HNC 3–2 toward the protoplanetary disks around the T Tauri star TW Hya and the Herbig Ae star HD 163296. HNC is detected toward both disks, constituting the first spatially resolved observations of HNC in disks. We also present SMA observations of HCN 3–2 and IRAM 30 m observations of HCN and HNC 1–0 toward HD 163296. The disk-averaged HNC/HCN emission ratio is 0.1–0.2 toward both disks. Toward TW Hya, the HNC emission is confined to a ring. The varying HNC abundance in the TW Hya disk demonstrates that HNC chemistry is strongly linked to the disk physical structure. In particular, the inner rim of the HNC ring can be explained by efficient destruction of HNC at elevated temperatures, similar to what is observed in the ISM. However, to realize the full potential of HNC as a disk tracer requires a combination of high SNR spatially resolved observations of HNC and HCN and disk-specific HNC chemical modeling.

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)

Plastic Flow Characteristics of Uranium-Niobium as a Function of Strain Rate and Temperature

International Nuclear Information System (INIS)

Cady, C.M.; Gray, G.T. III; Hecker, S.S; Thoma, D.J.; Korzekwa, D.R.; Patterson, R.A.; Dunn, P.S.; Bingert, J.F.

1999-01-01

The stress-strain response of uranium-niobium alloys as a function of temperature, strain-rate and stress-state was investigated. The yield and flow stresses of the U-Nb alloys were found to exhibit a pronounced strain rate sensitivity, while the hardening rates were found to be insensitive to strain rate and temperature. The overall stress-strain response of the U-6Nb exhibits a sinusoidal hardening response, which is consistent with multiple deformation modes and is thought to be related to shape-memory behavior

Study on microstructure change of Uranium nitride coated U-7wt%Mo powder by heat treatment

Energy Technology Data Exchange (ETDEWEB)

Cho, Woo Hyoung; Park, Jae Soon; Lee, Hae In; Kim, Woo Jeong; Yang, Jae Ho; Park, Jong Man [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2011-10-15

Uranium-molybdenum alloy particle dispersion fuel in an aluminum matrix with a high uranium density has been developed for a high performance research reactor in the RERTR program. In order to retard the fuel-matrix interaction in U-Mo/Al dispersion fuel in which the U-Mo fuel particles were dispersed in Al matrix, nitride layer coated U-Mo fuel particle has been designed and techniques to fabricate nitride-layer coated U-7wt%Mo particles have been developed in our lab. In this study, uranium nitride coated U-Mo particle has heat treatment for several times and degree. And we suggested for interaction layer remedy in U-Mo dispersion fuel. We investigate effect of heat treatment interaction layer evolution on uranium nitride coated U-Mo powder. The EDS and XRD analysis to investigate the phase evolution in uranium nitride coated layer is also a part of the present work

Fabrication of high-uranium-loaded U/sub 3/O/sub 8/-Al developmental fuel plates

Energy Technology Data Exchange (ETDEWEB)

Copeland, G.L.; Martin, M.M.

1980-12-01

A common plate-type fuel for research and test reactors is U/sub 3/O/sub 8/ dispersed in aluminum and clad with an aluminum alloy. There is an impetus to reduce the /sup 235/U enrichment from above 90% to below 20% for these fuels to lessen the risk of diversion of the uranium for nonpeaceful uses. Thus, the uranium content of the fuel plates has to be increased to maintain the performance of the reactors. This paper describes work at ORNL to determine the maximal uranium loading for these fuels that can be fabricated with commercially proven materials and techniques and that can be expected to perform satisfactorily in service.

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.

Development of Disk Rover, wall-climbing robot using permanent magnet disk

International Nuclear Information System (INIS)

Hirose, Shigeo; Tsutsumitake; Hiroshi; Toyama, Ryousei; Kobayashi, Kengo.

1992-01-01

A new type of wall climbing robot, named Disk Rover, using permanent magnet disks are developed. The newly introduced permanent magnet disk is to rotate the magnet disk on the surface of wall with partly contacted posture. It allows to produce high magnetic attraction force compared with conventional permanent wheel which utilizes only a small portion of the magnet installed around the wheel. The optimum design of the magnetic wheel is done by using finit element method and it is shown that the magnetic attraction force vs. weight ratio can be designed about three times higher than conventional type magnet wheel. The developed Disk Rover is 25 kg in weight including controller and battery, about 685 mm in diameter, 239 mm in height and has a pair of permanent magnet disks. It is demonstrated by the experiments that the Disk Rover can move around on the surface of the wall quite smoothly by radio control and has payload of about its own weight. Several considerations are also done in order to surmount bead weld. (author)

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)

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

New Fuel Alloys Seeking Optimal Solidus and Phase Behavior for High Burnup and TRU Burning

International Nuclear Information System (INIS)

Mariani, R.D.; Porter, D.L.; Kennedy, J.R.; Hayes, S.L.; Blackwood, V.S.; Jones, Z.S.; Olson, D.L.; Mishra, B.

2015-01-01

Recent modifications to fast reactor metallic fuels have been directed toward improving the melting and phase behaviors of the fuel alloy, for the purpose of ultra-high burnup and transuranic (TRU) burning. Improved melting temperatures increase the safety margin for uranium-based fast reactor fuel alloys, which is especially important for transuranic burning because the introduction of plutonium and neptunium acts to lower the alloy melting temperature. Improved phase behavior—single-phase, body-centered cubic—is desired because the phase is isotropic and the alloy properties are more predictable. An optimal alloy with both improvements was therefore sought through a comprehensive literature survey and theoretical analyses, and the creation and testing of some alloys selected by the analyses. Summarized here are those analyses, the impact of alloy modifications, and recent experimental results for selected pseudo-binary alloy systems that are hoped to accomplish the goals in a short timeframe. (author)

Analysis of heterogeneities in strain and microstructure in aluminum alloy and magnesium processed by high-pressure torsion

Energy Technology Data Exchange (ETDEWEB)

Panda, Subrata, E-mail: subrata.panda@univ-lorraine.fr [Université de Lorraine, Laboratory of Excellence on Design of Alloy Metals for low-mass Structures (DAMAS), Ile du Saulcy, Metz F-57045 (France); Université de Lorraine, Laboratoire d' Etude des Microstructures et de Mécanique des Matériaux (LEM3 UMR 7239), Ile du Saulcy, Metz F-57045 (France); Toth, Laszlo S., E-mail: laszlo.toth@univ-lorraine.fr [Université de Lorraine, Laboratory of Excellence on Design of Alloy Metals for low-mass Structures (DAMAS), Ile du Saulcy, Metz F-57045 (France); Université de Lorraine, Laboratoire d' Etude des Microstructures et de Mécanique des Matériaux (LEM3 UMR 7239), Ile du Saulcy, Metz F-57045 (France); Fundenberger, Jean-Jacques, E-mail: jean-jacques.fundenberger@univ-lorraine.fr [Université de Lorraine, Laboratory of Excellence on Design of Alloy Metals for low-mass Structures (DAMAS), Ile du Saulcy, Metz F-57045 (France); Université de Lorraine, Laboratoire d' Etude des Microstructures et de Mécanique des Matériaux (LEM3 UMR 7239), Ile du Saulcy, Metz F-57045 (France); Perroud, Olivier, E-mail: olivier.perroud@univ-lorraine.fr [Université de Lorraine, Laboratory of Excellence on Design of Alloy Metals for low-mass Structures (DAMAS), Ile du Saulcy, Metz F-57045 (France); Université de Lorraine, Laboratoire d' Etude des Microstructures et de Mécanique des Matériaux (LEM3 UMR 7239), Ile du Saulcy, Metz F-57045 (France); and others

2017-01-15

Two distinct bulk light metals were opted to study the shear strain evolution and associated heterogeneities in texture/microstructure development during torsional straining by high pressure torsion (HPT): a face centered cubic Al alloy (A5086) and a hexagonal commercial purity Mg. Relatively thick disk samples - four times thicker than usually employed in HPT process - were processed to 180° and 270° rotations. With the help of X-ray tomography, the shear strain gradients were examined in the axial direction. The results showed strongly localized shear deformation in the middle plane of the disks in both materials. These gradients involved strong heterogeneities in texture, microstructure and associated hardness, in particular through the thickness direction at the periphery of the disk where the interplay between significant strain hardening and possible dynamic recrystallization could occur. - Highlights: •HPT processing was conducted on bulk specimens thicker than the usual thin-disks. •The Al alloy (A5086) and commercial purity magnesium samples were compared. •Distributions of strain and microhardness were evaluated in the radial and axial direction. •Plastic deformation is highly localized in the middle plane at outer edge in both materials. •Different DRX rates governed the differences in microstructure and hardening behavior.

Review of the early AP penetrator work at LASL which led to the selection of U-3/4 Ti alloy

International Nuclear Information System (INIS)

Sandstrom, D.J.

1976-01-01

A historical review is presented of the depleted uranium penetrator work. The following alloys were studied: U--Ti, U--Mo, U--Nb, and U--Nb--Ti. Ballistic properties, armor penetration, and corrosion resistance were studied. The U--Ti alloy was found to offer the best combination of properties. 12 figures

Assessment of Low Cycle Fatigue Behavior of Powder Metallurgy Alloy U720

Science.gov (United States)

Gabb, Tomothy P.; Bonacuse, Peter J.; Ghosn, Louis J.; Sweeney, Joseph W.; Chatterjee, Amit; Green, Kenneth A.

2000-01-01

The fatigue lives of modem powder metallurgy disk alloys are influenced by variabilities in alloy microstructure and mechanical properties. These properties can vary as functions of variables the different steps of materials/component processing: powder atomization, consolidation, extrusion, forging, heat treating, and machining. It is important to understand the relationship between the statistical variations in life and these variables, as well as the change in life distribution due to changes in fatigue loading conditions. The objective of this study was to investigate these relationships in a nickel-base disk superalloy, U720, produced using powder metallurgy processing. Multiple strain-controlled fatigue tests were performed at 538 C (1000 F) at limited sets of test conditions. Analyses were performed to: (1) assess variations of microstructure, mechanical properties, and LCF failure initiation sites as functions of disk processing and loading conditions; and (2) compare mean and minimum fatigue life predictions using different approaches for modeling the data from assorted test conditions. Significant variations in life were observed as functions of the disk processing variables evaluated. However, the lives of all specimens could still be combined and modeled together. The failure initiation sites for tests performed at a strain ratio R(sub epsilon) = epsilon(sub min)/epsilon(sub max) of 0 were different from those in tests at a strain ratio of -1. An approach could still be applied to account for the differences in mean and maximum stresses and strains. This allowed the data in tests of various conditions to be combined for more robust statistical estimates of mean and minimum lives.

Fission and activation of uranium by fashion-plasma neutrons

International Nuclear Information System (INIS)

Lee, J.H.; Hochl, F.; McFarland, D.R.

1978-01-01

Disks of enriched and depleted uranium were irradiated by neutrons from the D-D fusions in a dense plasma-focus. A fission yield of 10 6 fissions-cm -3 in U 235 per pulse was determined with Ge(Li) gamme-ray spectrometry. Activation of U 238 caused increased beta activity after the plasma-neutron irradiation but alpha-particle spectrometry showed Pu 239 production was negligible. In addition, with a disk of lithium in the apparatus, 13.3 MeV neutrons from 7 Li(d,n) 8 Be was observed with a 80-m time-of-flight neutron detector. Dense plasma focuses are now operated not only in a single coaxial gun, but also in improved geometries, such as the hypocycloidal pinch and the staged plasma focus, from which a multiple plasma-focus array suitable for experimental verification of, and eventuel development into a fusion-fission hybrid reactor could be produced. (orig.) [de

Development of the white cast iron with niobium alloy, heat treating, to wear of the abrasive resistance

International Nuclear Information System (INIS)

Farah, Alessandro Fraga

1997-01-01

This work presents the heat treatment and abrasion tests results of a white cast iron with niobium alloy. The hardening heat treatment were made 950, 1000, 1050 e 110 deg C temperatures cooled by forced air. The tempering treatment were made at 450, 500 e 550 deg C temperatures. The heat treating alloy were compared, in the abrasive tests, with commercial alloys used as hardfacing by welding process in wear pieces. The abrasion tests was realized in pin on disk test. Additional tests were carried out for microstructural characterization to identify the different phases presents in the alloys. In a general way, the alloy studies showed the best wear rate for the heat treatments that results in higher hardness. It performance was superior than that of the commercial alloys. (author)

Understanding Floppy Disks.

Science.gov (United States)

Valentine, Pamela

1980-01-01

The author describes the floppy disk with an analogy to the phonograph record, and discusses the advantages, disadvantages, and capabilities of hard-sectored and soft-sectored floppy disks. She concludes that, at present, the floppy disk will continue to be the primary choice of personal computer manufacturers and their customers. (KC)

OT1_ipascucc_1: Understanding the Origin of Transition Disks via Disk Mass Measurements

Science.gov (United States)

Pascucci, I.

2010-07-01

Transition disks are a distinguished group of few Myr-old systems caught in the phase of dispersing their inner dust disk. Three different processes have been proposed to explain this inside-out clearing: grain growth, photoevaporation driven by the central star, and dynamical clearing by a forming giant planet. Which of these processes lead to a transition disk? Distinguishing between them requires the combined knowledge of stellar accretion rates and disk masses. We propose here to use 43.8 hours of PACS spectroscopy to detect the [OI] 63 micron emission line from a sample of 21 well-known transition disks with measured mass accretion rates. We will use this line, in combination with ancillary CO millimeter lines, to measure their gas disk mass. Because gas dominates the mass of protoplanetary disks our approach and choice of lines will enable us to trace the bulk of the disk mass that resides beyond tens of AU from young stars. Our program will quadruple the number of transition disks currently observed with Herschel in this setting and for which disk masses can be measured. We will then place the transition and the ~100 classical/non-transition disks of similar age (from the Herschel KP "Gas in Protoplanetary Systems") in the mass accretion rate-disk mass diagram with two main goals: 1) reveal which gaps have been created by grain growth, photoevaporation, or giant planet formation and 2) from the statistics, determine the main disk dispersal mechanism leading to a transition disk.

Defect microstructure in copper alloys irradiated with 750 MeV protons

DEFF Research Database (Denmark)

Zinkle, S.J.; Horsewell, A.; Singh, B.N.

1994-01-01

Transmission electron microscopy (TEM) disks of pure copper and solid solution copper alloys containing 5 at% of Al, Mn, or Ni were irradiated with 750 MeV protons to damage levels between 0.4 and 2 displacements per atom (dpa) at irradiation temperatures between 60 and 200 degrees C. The defect...... significant effect on the total density of small defect clusters, but they did cause a significant decrease in the fraction of defect clusters resolvable as SFT to similar to 20 to 25%. In addition, the dislocation loop density (> 5 nm diameter) was more than an order of magnitude higher in the alloys...

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)

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)

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

Exploring Disks Around Planets

Science.gov (United States)

Kohler, Susanna

2017-07-01

Giant planets are thought to form in circumstellar disks surrounding young stars, but material may also accrete into a smaller disk around the planet. Weve never detected one of these circumplanetary disks before but thanks to new simulations, we now have a better idea of what to look for.Image from previous work simulating a Jupiter-mass planet forming inside a circumstellar disk. The planet has its own circumplanetary disk of accreted material. [Frdric Masset]Elusive DisksIn the formation of giant planets, we think the final phase consists of accretion onto the planet from a disk that surrounds it. This circumplanetary disk is important to understand, since it both regulates the late gas accretion and forms the birthplace of future satellites of the planet.Weve yet to detect a circumplanetary disk thus far, because the resolution needed to spot one has been out of reach. Now, however, were entering an era where the disk and its kinematics may be observable with high-powered telescopes (like the Atacama Large Millimeter Array).To prepare for such observations, we need models that predict the basic characteristics of these disks like the mass, temperature, and kinematic properties. Now a researcher at the ETH Zrich Institute for Astronomy in Switzerland, Judit Szulgyi, has worked toward this goal.Simulating CoolingSzulgyi performs a series of 3D global radiative hydrodynamic simulations of 1, 3, 5, and 10 Jupiter-mass (MJ) giant planets and their surrounding circumplanetary disks, embedded within the larger circumstellar disk around the central star.Density (left column), temperature (center), and normalized angular momentum (right) for a 1 MJ planet over temperatures cooling from 10,000 K (top) to 1,000 K (bottom). At high temperatures, a spherical circumplanetary envelope surrounds the planet, but as the planet cools, the envelope transitions around 64,000 K to a flattened disk. [Szulgyi 2017]This work explores the effects of different planet temperatures and

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.

The development of uranium foil farication technology utilizing twin roll method for Mo-99 irradiation target

CERN Document Server

Kim, C K; Park, H D

2002-01-01

MDS Nordion in Canada, occupying about 75% of global supply of Mo-99 isotope, has provided the irradiation target of Mo-99 using the rod-type UAl sub x alloys with HEU(High Enrichment Uranium). ANL (Argonne National Laboratory) through co-operation with BATAN in Indonesia, leading RERTR (Reduced Enrichment for Research and Test Reactors) program substantially for nuclear non-proliferation, has designed and fabricated the annular cylinder of uranium targets, and successfully performed irradiation test, in order to develop the fabrication technology of fission Mo-99 using LEU(Low Enrichment Uranium). As the uranium foils could be fabricated in laboratory scale, not in commercialized scale by hot rolling method due to significant problems in foil quality, productivity and economic efficiency, attention has shifted to the development of new technology. Under these circumstances, the invention of uranium foil fabrication technology utilizing twin-roll casting method in KAERI is found to be able to fabricate LEU or...

Cytocompatibility of a free machining titanium alloy containing lanthanum.

Science.gov (United States)

Feyerabend, Frank; Siemers, Carsten; Willumeit, Regine; Rösler, Joachim

2009-09-01

Titanium alloys like Ti6Al4V are widely used in medical engineering. However, the mechanical and chemical properties of titanium alloys lead to poor machinability, resulting in high production costs of medical products. To improve the machinability of Ti6Al4V, 0.9% of the rare earth element lanthanum (La) was added. The microstructure, the mechanical, and the corrosion properties were determined. Lanthanum containing alloys exhibited discrete particles of cubic lanthanum. The mechanical properties and corrosion resistance were slightly decreased but are still sufficient for many applications in the field of medical engineering. In vitro experiments with mouse macrophages (RAW 264.7) and human bone-derived cells (MG-63, HBDC) were performed and revealed that macrophages showed a dose response below and above a LaCl3 concentration of 200 microM, while MG-63 and HBDC tolerated three times higher concentrations without reduction of viability. The viability of cells cultured on disks of the materials showed no differences between the reference and the lanthanum containing alloy. We therefore propose that lanthanum containing alloy appears to be a good alternative for biomedical applications, where machining of parts is necessary.

Corrosion of high Ni-Cr alloys and Type 304L stainless steel in HNO3-HF

International Nuclear Information System (INIS)

Ondrejcin, R.S.; McLaughlin, B.D.

1980-04-01

Nineteen alloys were evaluated as possible materials of construction for steam heating coils, the dissolver vessel, and the off-gas system of proposed facilities to process thorium and uranium fuels. Commercially available alloys were found that are satisfactory for all applications. With thorium fuel, which requires HNO 3 -HF for dissolution, the best alloy for service at 130 0 C when complexing agents for fluoride are used is Inconel 690; with no complexing agents at 130 0 C, Inconel 671 is best. At 95 0 C, six other alloys tested would be adequate: Haynes 25, Ferralium, Inconel 625, Type 304L stainless steel, Incoloy 825, and Haynes 20 (in order of decreasing preference); based on composition, six untested alloys would also be adequate. The ions most effective in reducing fluoride corrosion were the complexing agents Zr 4+ and Th 4+ ; Al 3+ was less effective. With uranium fuel, modestly priced Type 304L stainless steel is adequate. Corrosion will be most severe in HNO 3 -HF used occasionally for flushing and in solutions of HNO 3 and corrosion products (ferric and dichromate ions). HF corrosion can be minimized by complexing the fluoride ion and by passivation of the steel with strong nitric acid. Corrosion caused by corrosion products can be minimized by operating at lower temperatures

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.

Tungsten versus depleted uranium for armour-piercing penetrators

International Nuclear Information System (INIS)

Johnson, P.K.

1983-01-01

Tungsten alloys have been widely used in the production of armour-piercing (AP) penetrators for defense purposes for the past 40 years. In recent years, however, depleted uranium (DU) has also been utilised for this application. Both materials exhibit high density and strength, two properties necessary for kinetic-energy projectiles to penetrate armour on tanks and other vehicles. The facts, however, support the view that tungsten can and should be utilised as the primary material for most armour-defeating ordnance applications. (author)

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

Tribological Characteristic of Titanium Alloy Surface Layers Produced by Diode Laser Gas Nitriding

Directory of Open Access Journals (Sweden)

Lisiecki A.

2016-06-01

Full Text Available In order to improve the tribological properties of titanium alloy Ti6Al4V composite surface layers Ti/TiN were produced during laser surface gas nitriding by means of a novel high power direct diode laser with unique characteristics of the laser beam and a rectangular beam spot. Microstructure, surface topography and microhardness distribution across the surface layers were analyzed. Ball-on-disk tests were performed to evaluate and compare the wear and friction characteristics of surface layers nitrided at different process parameters, base metal of titanium alloy Ti6Al4V and also the commercially pure titanium. Results showed that under dry sliding condition the commercially pure titanium samples have the highest coefficient of friction about 0.45, compared to 0.36 of titanium alloy Ti6Al4V and 0.1-0.13 in a case of the laser gas nitrided surface layers. The volume loss of Ti6Al4V samples under such conditions is twice lower than in a case of pure titanium. On the other hand the composite surface layer characterized by the highest wear resistance showed almost 21 times lower volume loss during the ball-on-disk test, compared to Ti6Al4V samples.

Equilibrium configuration of a stratus floating above accretion disks: Full-disk calculation

Science.gov (United States)

Itanishi, Yusuke; Fukue, Jun

2017-06-01

We examine floating strati above a luminous accretion disk, supported by the radiative force from the entire disk, and calculate the equilibrium locus, which depends on the disk luminosity and the optical depth of the stratus. Due to the radiative transfer effect (albedo effect), the floating height of the stratus with a finite optical depth generally becomes high, compared with the particle case. In contrast to the case of the near-disk approximation, moreover, the floating height becomes yet higher in the present full-disk calculation, since the intense radiation from the inner disk is taken into account. As a result, when the disk luminosity normalized by the Eddington luminosity is ˜0.3 and the stratus optical depth is around unity, the stable configuration disappears at around r ˜ 50 rg, rg being the Schwarzschild radius, and the stratus would be blown off as a cloudy wind consisting of many strati with appropriate conditions. This luminosity is sufficiently smaller than the Eddington one, and the present results suggest that the radiation-driven cloudy wind can be easily blown off from the sub-Eddington disk, and this can explain various outflows observed in ultra-fast outflow objects as well as in broad-absorption-line quasars.

New Fuel Alloys Seeking Optimal Solidus and Phase Behavior for High Burnup and TRU Burning

International Nuclear Information System (INIS)

Blackwood, V.S.; Jones, Z.S.; Olson, D.L.; Mishra, B.; Mariani, R.D.; Porter, D.L.; Kennedy, J.R.; Hayes, S.L.

2013-01-01

Summary: • Pd will bind lanthanide fission products. • 2 wt% Pd in alloy is expected to allow 20 at% Heavy Metal burnup, 4 wt% Pd possibly 30-40 at% HM burnup. • For recycled fuel with some lanthanide carryover, palladium additive will also prevent premature FCCI. • Novel uranium alloy systems suitable for burning transuranics were identified. • U-Mo-Ti-Zr and U-W-Mo irradiations may perform comparably to U-10Zr, but the real tests needed must include Pu and Np for TRU burning. – Diffusion couples with alloys and Fe or cladding; – Irradiations

NEW SERDP Project: Copper- Beryllium Alternatives Alloys Development

Science.gov (United States)

2011-02-10

Nitronic60, HBN 304 stainless steel , as well as low friction coating\\liner systems on PH stainless steel substrates • Compression strength and...ChemistryRefining Lath Martensite : Ms≥200°C Nickel: Cleavage Resistance Cobalt: SRO Recovery Resistance Chromium: Corrosion Resistance σuts > 280 ksi σys...against representative steels ). o Compression testing from each of the Cu- and Co-based alloys will be performed per ASTM E 9 o Pin-on-Disk test per

Obtaining of U-2.5Zr7.5Nb and U-3Zr-9Nb alloys by sintering process

International Nuclear Information System (INIS)

Mazzeu, Thiago de Oliveira; Paula, Joao Bosco de; Ferraz, Wilmar Barbosa; Santos, Ana Maria Matildes dos; Brina, Jose Giovanni Mascarenhas

2011-01-01

The development of metallic fuels with low enrichment to be used in research and test reactors, as well in the future pressurized water reactors, focuses on the search for uranium alloys of high density. Alloying elements such as Zr, Nb and Mo are added to uranium to improve fuel performance in reactors. In this context, the Centro de Desenvolvimento da Tecnologia Nuclear (CDTN) in Belo Horizonte is developing the U-2.5Zr-7.5Nb and U- 3Zr-9Nb (weight %) alloys by the innovative process of sintering that utilizes raw materials in the form of powders. The powders were pressed at 400MPa and then sintered under a vacuum of about 5 x 10-6 Torr at temperatures ranging from 1050 deg to 1300 deg C. The densities of the alloys were measured geometrically and by hydrostatic method using water. The microstructures of the pellets were observed by scanning electron microscopy (SEM) and the elements of alloying were identified by energy dispersive X-ray spectroscopy (SEM/EDS) analysis. The obtained results showed a small increasing density with rising sintering temperature. The highest density achieved was approximately 80% of theoretical density. It was also qualitatively observed that the superficial oxidation of the pellets increased with increasing sintering temperature thus avoiding the fusion of the alloys at higher temperatures. (author)

Benzotriazole as a passivating agent during chemical mechanical planarization of Ni–P alloy substrates

International Nuclear Information System (INIS)

Mu, Yan; Zhong, Mingjie; Rushing, Kenneth J.; Li, Yuzhuo; Shipp, Devon A.

2014-01-01

Highlights: • Benzotriazole (BTA) is used to passivate the Chemical Mechanical Planarization of Ni-P alloys. • BTA significantly decreases the average R a of the polished surfaces at low concentrations. • XPS, AFM and electrochemical studies are used to probe passivation effects of BTA on Ni–P surfaces. • Findings potentially impact hard disk drive manufacturing processes. - Abstract: With the rapid increase of data storage density on computer hard disk drives (HDDs), the operation distance between read/write head and disk surface has fallen to just a few nanometers. Chemical mechanical planarization (CMP) has been selected as the best process to produce high quality surface finish during the manufacturing of Ni–P alloy substrates for HDD applications. Herein we report, for the first time, the use of benzotriazole (BTA) as a passivating agent in CMP slurries to decrease the surface roughness (R a ). Results show that the average R a of the polished surfaces is decreased to 0.2 nm in a 5 μm × 5 μm scan area with the adding of 2 mM BTA. X-ray photoelectron spectroscopy (XPS) and electrochemical studies results further prove the interaction between BTA and Ni–P surface and the formation of an effective passivating layer on Cu in CMP slurries containing BTA

Powder fabrication of U-Mo alloys for nuclear dispersion fuels

Energy Technology Data Exchange (ETDEWEB)

Durazzo, Michelangelo; Rocha, Claudio Jose da; Mestnik Filho, Jose; Leal Neto, Ricardo Mendes, E-mail: mdurazzo@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

2011-07-01

For the last 30 years high uranium density dispersion fuels have been developed in order to accomplish the low enrichment goals of the Reduced Enrichment for Research and Test Reactors (RERTR) Program. Gamma U-Mo alloys, particularly with 7 to 10 wt% Mo, as a fuel phase dispersed in aluminum matrix, have shown good results concerning its performance under irradiation tests. That's why this fissile phase is considered to be used in the nuclear fuel of the Brazilian Multipurpose Research Reactor (RMB), currently being designed. Powder production from these ductile alloys has been attained by atomization, mechanical (machining, grinding, cryogenic milling) and chemical (hydriding-de hydriding) methods. This work is a part of the efforts presently under way at IPEN to investigate the feasibility of these methods. Results on alloy fabrication by induction melting and gamma-stabilization of U-10Mo alloys are presented. Some results on powder production and characterization are also discussed. (author)

Powder fabrication of U-Mo alloys for nuclear dispersion fuels

International Nuclear Information System (INIS)

Durazzo, Michelangelo; Rocha, Claudio Jose da; Mestnik Filho, Jose; Leal Neto, Ricardo Mendes

2011-01-01

For the last 30 years high uranium density dispersion fuels have been developed in order to accomplish the low enrichment goals of the Reduced Enrichment for Research and Test Reactors (RERTR) Program. Gamma U-Mo alloys, particularly with 7 to 10 wt% Mo, as a fuel phase dispersed in aluminum matrix, have shown good results concerning its performance under irradiation tests. That's why this fissile phase is considered to be used in the nuclear fuel of the Brazilian Multipurpose Research Reactor (RMB), currently being designed. Powder production from these ductile alloys has been attained by atomization, mechanical (machining, grinding, cryogenic milling) and chemical (hydriding-de hydriding) methods. This work is a part of the efforts presently under way at IPEN to investigate the feasibility of these methods. Results on alloy fabrication by induction melting and gamma-stabilization of U-10Mo alloys are presented. Some results on powder production and characterization are also discussed. (author)

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.

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.

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

Magnesium and uranium ignition in different gaseous atmospheres; Inflammabilite du magnesium et de l'uranium dans l'air et le gaz carbonique

Energy Technology Data Exchange (ETDEWEB)

Darras, R; Baque, P; Leclercq, D [Commissariat a l' Energie Atomique, Saclay (France).Centre d' Etudes Nucleaires

1960-07-01

Magnesium, uranium and some of their alloys burning temperatures have been systematically determined in an air or carbon dioxide atmosphere, either dry or wet. Two different ways of heating have been used: either continuously rising up the temperature, or heating to and then maintaining a constant temperature. The results are clearly different in the two cases. Besides, if moisture has little effect on the magnesium burning temperatures in air, it does lower them by about 130-140 deg. C in CO{sub 2}. The differences of sight between the burning of magnesium and uranium have been noticed; this leads to distinguish between an 'ignition' and an 'inflammation'. (author) [French] Les temperatures auxquelles apparait la combustion vive du magnesium, de l'uranium et certains de leurs alliages ont ete determinees systematiquement dans l'air et le gaz carbonique, soit secs, soit humidifies. On a mis en evidence l'influence du mode de chauffage sur les resultats: soit montee en temperature continue, soit stabilisation a partir d'une certaine temperature. En outre, si la presence d'humidite affecte peu les temperatures de combustion vive du magnesium dans l'air, elle les abaisse de 130 a 140 deg. C dans le gaz carbonique. Les differences d'aspect entre la combustion vive du magnesium et de l'uranium ont egalement ete remarquees, ce qui amene notamment a distinguer une 'ignition' d'une 'inflammation'. (auteur)

2TB hard disk drive

CERN Multimedia

This particular object was used up until 2012 in the Data Centre. It slots into one of the Disk Server trays. Hard disks were invented in the 1950s. They started as large disks up to 20 inches in diameter holding just a few megabytes (link is external). They were originally called "fixed disks" or "Winchesters" (a code name used for a popular IBM product). They later became known as "hard disks" to distinguish them from "floppy disks (link is external)." Hard disks have a hard platter that holds the magnetic medium, as opposed to the flexible plastic film found in tapes and floppies.

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

Distribution of uranium in marine sediments; Distribucion de uranio en sedimentos marinos

Energy Technology Data Exchange (ETDEWEB)

Ordonez R, E.; Ramirez T, J.J.; Lopez M, J.; Aspiazu, J. [ININ, 52750 La Marquesa, Estado de Mexico (Mexico); Ruiz F, A.C. [U. Academica Mazatlan, ICML, UNAM (Mexico); Valero C, N. [CONALEP, 52000 Lerma, Estado de Mexico (Mexico)

2008-07-01

The marine sediments obtained by means of a sampling nucleus in the Gulf of Tehuantepec, Mexico, they have been object of crystallographic and morphological characterization. The PIXE analysis of some samples in study is shown. The normal methodology to carry out the alpha spectroscopy indicates that the sample should be dissolved, but due to the nature of the marine sediments, it thinks about the necessity to make a fractional separation of the sample components. In each stratum of the profile it separates the organic part and the mineral to recover the uranium. It was observed that in the organic phase, the uranium is in two oxidation states (IV and Vl), being necessary the radiochemical separation with a liquid/liquid column chromatographic that uses the di-2-ethyl hexyl phosphoric acid as stationary phase. The uranium compounds extracts are electrodeposited in fine layers on stainless steel disks to carry out the analysis by alpha spectroscopy. The spectroscopic analysis of the uranium indicates us that for each stratum one has a difference marked in the quotient of activities of {sup 234}U/{sup 238}U that depends on the nature of the studied fraction. These results give us a clear idea about how it is presented the effect of the uranium migration and other radioelements in the biosphere, with what we can determine which are the conditions in that these have their maximum mobility and to know their diffusion patterns in the different media studied. (Author)

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)

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)

DISK DETECTIVE: DISCOVERY OF NEW CIRCUMSTELLAR DISK CANDIDATES THROUGH CITIZEN SCIENCE

Energy Technology Data Exchange (ETDEWEB)

Kuchner, Marc J.; McElwain, Michael; Padgett, Deborah L. [NASA Goddard Space Flight Center Exoplanets and Stellar Astrophysics Laboratory, Code 667 Greenbelt, MD 21230 (United States); Silverberg, Steven M.; Wisniewski, John P. [Homer L. Dodge Department of Physics and Astronomy The University of Oklahoma 440 W. Brooks St. Norman, OK 73019 (United States); Bans, Alissa S. [Valparaiso University, Department of Physics and Astronomy, Neils Science Center, 1610 Campus Drive East, Valparaiso, IN 46383 (United States); Bhattacharjee, Shambo [International Space University 1 Rue Jean-Dominique Cassini F-67400 Illkirch-Graffenstaden (France); Kenyon, Scott J. [Smithsonian Astrophysical Observatory 60 Garden Street Cambridge, MA 02138 (United States); Debes, John H. [Space Telescope Science Institute 3700 San Martin Dr. Baltimore, MD 21218 (United States); Currie, Thayne [National Astronomical Observatory of Japan 650 N A’ohokhu Place Hilo, HI 96720 (United States); García, Luciano [Observatorio Astronómico de Córdoba Universidad Nacional de Córdoba Laprida 854, X5000BGR, Córdoba (Argentina); Jung, Dawoon [Korea Aerospace Research Institute Lunar Exploration Program Office 169-84 Gwahak-ro, Yuseong-gu, Daejeon 34133 (Korea, Republic of); Lintott, Chris [Denys Wilkinson Building Keble Road Oxford, OX1 3RH (United Kingdom); Rebull, Luisa M. [Infrared Processing and Analaysis Center Caltech M/S 314-6 1200 E. California Blvd. Pasadena, CA 91125 (United States); Nesvold, Erika, E-mail: Marc.Kuchner@nasa.gov, E-mail: michael.w.mcelwain@nasa.gov, E-mail: deborah.l.padgett@nasa.gov, E-mail: carol.a.grady@nasa.gov, E-mail: silverberg@ou.edu, E-mail: wisniewski@ou.edu [Department of Terrestrial Magnetism, Carnegie Institution of Washington, 5241 Broad Branch Road, NW, Washington, DC 20015-1305 (United States); Collaboration: Disk Detective Collaboration; and others

2016-10-20

The Disk Detective citizen science project aims to find new stars with 22 μ m excess emission from circumstellar dust using data from NASA’s Wide-field Infrared Survey Explorer ( WISE ) mission. Initial cuts on the AllWISE catalog provide an input catalog of 277,686 sources. Volunteers then view images of each source online in 10 different bands to identify false positives (galaxies, interstellar matter, image artifacts, etc.). Sources that survive this online vetting are followed up with spectroscopy on the FLWO Tillinghast telescope. This approach should allow us to unleash the full potential of WISE for finding new debris disks and protoplanetary disks. We announce a first list of 37 new disk candidates discovered by the project, and we describe our vetting and follow-up process. One of these systems appears to contain the first debris disk discovered around a star with a white dwarf companion: HD 74389. We also report four newly discovered classical Be stars (HD 6612, HD 7406, HD 164137, and HD 218546) and a new detection of 22 μ m excess around the previously known debris disk host star HD 22128.

DISK DETECTIVE: DISCOVERY OF NEW CIRCUMSTELLAR DISK CANDIDATES THROUGH CITIZEN SCIENCE

International Nuclear Information System (INIS)

Kuchner, Marc J.; McElwain, Michael; Padgett, Deborah L.; Silverberg, Steven M.; Wisniewski, John P.; Bans, Alissa S.; Bhattacharjee, Shambo; Kenyon, Scott J.; Debes, John H.; Currie, Thayne; García, Luciano; Jung, Dawoon; Lintott, Chris; Rebull, Luisa M.; Nesvold, Erika

2016-01-01

The Disk Detective citizen science project aims to find new stars with 22 μ m excess emission from circumstellar dust using data from NASA’s Wide-field Infrared Survey Explorer ( WISE ) mission. Initial cuts on the AllWISE catalog provide an input catalog of 277,686 sources. Volunteers then view images of each source online in 10 different bands to identify false positives (galaxies, interstellar matter, image artifacts, etc.). Sources that survive this online vetting are followed up with spectroscopy on the FLWO Tillinghast telescope. This approach should allow us to unleash the full potential of WISE for finding new debris disks and protoplanetary disks. We announce a first list of 37 new disk candidates discovered by the project, and we describe our vetting and follow-up process. One of these systems appears to contain the first debris disk discovered around a star with a white dwarf companion: HD 74389. We also report four newly discovered classical Be stars (HD 6612, HD 7406, HD 164137, and HD 218546) and a new detection of 22 μ m excess around the previously known debris disk host star HD 22128.

Disk Detective: Discovery of New Circumstellar Disk Candidates Through Citizen Science

Science.gov (United States)

Kuchner, Marc J.; Silverberg, Steven M.; Bans, Alissa S.; Bhattacharjee, Shambo; Kenyon, Scott J.; Debes, John H.; Currie, Thayne; Garcia, Luciano; Jung, Dawoon; Lintott, Chris;

Continuing investigations for technology assessment of 99Mo production from LEU [low enriched uranium] targets

International Nuclear Information System (INIS)

Vandegrift, G.F.; Kwok, J.D.; Marshall, S.L.; Vissers, D.R.; Matos, J.E.

1987-01-01

Currently much of the world's supply of 99m Tc for medical purposes is produced from 99 Mo derived from the fissioning of high enriched uranium (HEU). This paper presents the results of our continuing studies on the effects of substituting low enriched uranium (LEU) for HEU in targets for the production of fission product 99 Mo. Improvements in the electrodeposition of thin films of uranium metal continue to increase the appeal for the substitution of LEU metal for HEU oxide films in cylindrical targets. The process is effective for targets fabricated from stainless steel or zircaloy. Included is a cost estimate for setting up the necessary equipment to electrodeposit uranium metal on cylindrical targets. Further investigations on the effect of LEU substitution on processing of these targets are also reported. Substitution of uranium silicides for the uranium-aluminium alloy or uranium aluminide dispersed fuel used in current target designs will allow the substitution of LEU for HEU in these targets with equivalent 99 Mo-yield per target and no change in target geometries. However, this substitution will require modifications in current processing steps due to 1) the insolubility of uranium silicides in alkaline solutions and 2) the presence of significant quantities of silicate in solution. Results to date suggest that substitution of LEU for HEU can be achieved. (Author)

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

Magnetohydrodynamics of accretion disks

International Nuclear Information System (INIS)

Torkelsson, U.

1994-04-01

The thesis consists of an introduction and summary, and five research papers. The introduction and summary provides the background in accretion disk physics and magnetohydrodynamics. The research papers describe numerical studies of magnetohydrodynamical processes in accretion disks. Paper 1 is a one-dimensional study of the effect of magnetic buoyancy on a flux tube in an accretion disk. The stabilizing influence of an accretion disk corona on the flux tube is demonstrated. Paper 2-4 present numerical simulations of mean-field dynamos in accretion disks. Paper 11 verifies the correctness of the numerical code by comparing linear models to previous work by other groups. The results are also extended to somewhat modified disk models. A transition from an oscillatory mode of negative parity for thick disks to a steady mode of even parity for thin disks is found. Preliminary results for nonlinear dynamos at very high dynamo numbers are also presented. Paper 3 describes the bifurcation behaviour of the nonlinear dynamos. For positive dynamo numbers it is found that the initial steady solution is replaced by an oscillatory solution of odd parity. For negative dynamo numbers the solution becomes chaotic at sufficiently high dynamo numbers. Paper 4 continues the studies of nonlinear dynamos, and it is demonstrated that a chaotic solution appears even for positive dynamo numbers, but that it returns to a steady solution of mixed parity at very high dynamo numbers. Paper 5 describes a first attempt at simulating the small-scale turbulence of an accretion disk in three dimensions. There is only find cases of decaying turbulence, but this is rather due to limitations of the simulations than that turbulence is really absent in accretion disks

Preparation of thin actinide metal disks using a multiple disk casting technique

International Nuclear Information System (INIS)

Conner, W.V.

1975-01-01

A casting technique has been developed for preparing multiple actinide metal disks which have a minimum thickness of 0.006 inch. This technique was based on an injection casting procedure which utilizes the weight of a tantalum metal rod to force the molten metal into the mold cavity. Using the proper mold design and casting parameters, it has been possible to prepare ten 1/2 inch diameter neptunium or plutonium metal disks in a single casting, This casting technique is capable of producing disks which are very uniform. The average thickness of the disks from a typical casting will vary no more than 0.001 inch and the variation in the thickness of the individual disks will range from 0.0001 to 0.0005 inch. (Auth.)

Preparation of thin actinide metal disks using a multiple disk casting technique

International Nuclear Information System (INIS)

Conner, W.V.

1976-01-01

A casting technique has been developed for preparing multiple actinide metal disks which have a minimum thickness of 0.006 inch. This technique was based on an injection casting procedure which utilizes the weight of a tantalum metal rod to force the molten metal into the mold cavity. Using the proper mold design and casting parameters, it has been possible to prepare ten 1/2 inch diameter neptunium or plutonium metal disks in a single casting. This casting technique is capable of producing disks which are very uniform. The average thickness of the disks from a typical casting will vary no more than 0.001 inch and the variation in the thickness of the individual disks will range from 0.0001 to 0.0005 inch. (author)

Separating the Influence of Environment from Stress Relaxation Effects on Dwell Fatigue Crack Growth in a Nickel-Base Disk Alloy

Science.gov (United States)

Telesman, J.; Gabb, T. P.; Ghosn, L. J.

2016-01-01

Both environmental embrittlement and crack tip visco-plastic stress relaxation play a significant role in determining the dwell fatigue crack growth (DFCG) resistance of nickel-based disk superalloys. In the current study performed on the Low Solvus High Refractory (LSHR) disk alloy, the influence of these two mechanisms were separated so that the effects of each could be quantified and modeled. Seven different microstructural variations of LSHR were produced by controlling the cooling rate and the subsequent aging and thermal exposure heat treatments. Through cyclic fatigue crack growth testing performed both in air and vacuum, it was established that four out of the seven LSHR heat treatments evaluated, possessed similar intrinsic environmental resistance to cyclic crack growth. For these four heat treatments, it was further shown that the large differences in dwell crack growth behavior which still persisted, were related to their measured stress relaxation behavior. The apparent differences in their dwell crack growth resistance were attributed to the inability of the standard linear elastic fracture mechanics (LEFM) stress intensity parameter to account for visco-plastic behavior. Crack tip stress relaxation controls the magnitude of the remaining local tensile stresses which are directly related to the measured dwell crack growth rates. It was hypothesized that the environmentally weakened grain boundary crack tip regions fail during the dwells when their strength is exceeded by the remaining local crack tip tensile stresses. It was shown that the classical creep crack growth mechanisms such as grain boundary sliding did not contribute to crack growth, but the local visco-plastic behavior still plays a very significant role by determining the crack tip tensile stress field which controls the dwell crack growth behavior. To account for the influence of the visco-plastic behavior on the crack tip stress field, an empirical modification to the LEFM stress

EDX microanalysis of neutron-irradiated alloys

International Nuclear Information System (INIS)

Thomas, L.E.

1981-09-01

Energy-dispersive X-ray (EDX) spectrometry of 50 nm thick specimens in the scanning transmission electron microscope provides quantitative elemental analyses of selected regions as small as 20 nm in diameter. To analyze highly radioactive neutron-irradiated alloys it is necessary to reduce the high counting deadtimes caused by energetic γ-Compton scattering in the Si(Li) detector, and to account for spurious background contributions from γ-rays and characteristic x-ray emissions. Several simple methods for overcoming effects of specimen radioactivity are described, including use of a tungsten collimator to attenuate γ and x-rays coming from the thick edges of self-supporting disk specimens. These methods allow analyses of Fe-Cr-Ni based alloys with γ-activities up to 1000 μC/sub i/. Techniques used to maintain high spatial resolution and accuracy in quantitatve analysis are also described, and their use is illustrated

Composition analysis of Ta-W alloy using NAA and EDXRF techniques

International Nuclear Information System (INIS)

Swain, K.K.; Remya Devi, P.S.; Chavan, Trupti A.; Verma, R.; Reddy, A.V.R.

2015-01-01

Tantalum-Tungsten (Ta-W) alloy is a high strength alloy and is used in corrosion resistant chemical process equipment's including heat exchangers, condensers, heating and cooling coils and reaction vessels. Ta-W alloy is also used as ion extraction plate during laser Isotope separation of uranium and hence the composition is critical for its optimal application. The composition of the alloy was determined by neutron activation analysis (NAA) and energy dispersive X-ray fluorescence spectrometry (EDXRF) techniques. Ta-W alloy sample was received from Nuclear Fuel Complex (NFC), Hyderabad. For NAA, samples (50 - 500 mg) were sealed in polyethylene. High purity Ta foil (30 - 40 mg) and W foil (10 - 20 mg) were packed and used as comparators. Samples and standards were irradiated in the graphite reflector position of Advanced Heavy Water Reactor Critical Facility (AHWR CF) reactor, BARC, Mumbai for 4 hours. After suitable decay period, radioactivity assay was carried out using a 45% relative efficiency high purity germanium (HPGe) detector coupled to MCA with 8 k conversion gain