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Density Prediction of Uranium-6 Niobium Ingots

Description: The densities of uranium-6 niobium (U-Nb) alloys have been compiled from a variety of literature sources such as Y-12 and Rocky Flats datasheets. We also took advantage of the 42 well-pedigreed, homogeneous baseline U-Nb alloys produced under the Enhanced Surveillance Program for density measurements. Even though U-Nb alloys undergo two-phase transitions as the Nb content varies from 0 wt. % to 8 wt %, the theoretical and measured densities vary linearly with Nb content. Therefore, the effect of Nb content on the density was modeled with a linear regression. From this linear regression, a homogeneous ingot of U-6 wt.% Nb would have a density of 17.382 {+-} 0.040 g/cc (95% CI). However, ingots produced at Y-12 are not homogeneous with respect to the Nb content. Therefore, using the 95% confidence intervals, the density of a Y-12 produced ingot would vary from 17.310 {+-} 0.043 g/cc at the center to 17.432 {+-} 0.039 g/cc at the edge. Ingots with larger Nb inhomogeneities will also have larger variances in the density.
Date: April 15, 2003
Creator: D.F.Teter; Tubesing, P.K.; D.J.Thoma & E.J.Peterson
Partner: UNT Libraries Government Documents Department

Preliminary investigation of grain refinement in a U-0.2 wt % V alloy casting by true isothermal transformation at 516{degree}C

Description: Laboratory-scale isothermal transformation from beta phase to alpha phase at 516 C was accomplished using one U-0.2 wt % V alloy composition and with specimens up to 7.6 mm in thickness. Gravity was used to transfer individual specimens from a furnace at 720 C to one at 516 C. The lower-temperature, furnace contained two copper blocks between which the specimens were quenched by contact. The furnace also contained a partial atmosphere of helium. Results duplicate those of Reisse et al. at this temperature. Their work was done on smaller samples. Grain sizes obtained were consistently ASTM 7 to 8 (20 to 30 {mu}m), indicating that the cooling rates at the center of even the thicker samples were adequate to miss the nose of the (upper) TTT curve. The microstructure obtained, including the grain size, appears to be equivalent to that obtained by carefully controlled wrought processing, but we believe these castings lack the strong crystallographic textures that exist in wrought products.
Date: July 1, 1995
Creator: Wood, D.H.; Flores, R. & Kershaw, R.P.
Partner: UNT Libraries Government Documents Department

Mechanical fabrication, heat treatment, and machining of uranium alloys

Description: From conference on physical metallurgy of uranium alloys; Vail, Colorado, USA (12 Feb 1974). A review of the state of the art is presented on the fabrication and properties of U-rich alloys in which the alloying elements are highly miscible in gamma -U (Nb, Zr, Ti, Mo). Processing of both low alloys (which will not retain a metastable gamma phase when quenched) and high alloys (which will retain a metastable gamma phase, usually greater than 6% Mo or Nb) is treated. Forging, extrusion, deep drawing, shear spinning, single-point turning, milling, drilling, tapping, and electrical discharge machining are discussed. Heat treatments (metastable, solution treatment, age hardening) to yield the desired mechanical properties are also discussed. (18 figures, 6 tables, 57 references) (DLC)
Date: January 1, 1974
Creator: Boland, J.F. & Sandstrom, D.J.
Partner: UNT Libraries Government Documents Department

The potential pyrophoricity of BMI-SPEC and aluminum plate spent fuels retrieved from underwater storage

Description: Physical/chemical factors in U metal and hydride combustion, particularly pyrophoricity in ambient environment, were evaluated for BMI-SPEC and UAl{sub x} plate fuels. Some metal fuels may be highly reactive (spontaneously igniting in air) due to high specific surface area, high decay heat, or a high U hydride content from corrosion during underwater storage. However, for the BMI-SPEC and the aluminum plate fuels, this reactivity is too low to present a realistic threat of uncontrolled spontaneous combustion at ambient conditions. While residual U hydride is expected in these corroded fuels, the hydride levels are expected to be too low and the configuration too unfavorable to ignite the fuel meat when the fuels are retrieved from the basin and dried. Furthermore the composition and microstructure of the UAl{sub x} fuels further mitigate that risk.
Date: August 1996
Creator: Ebner, M. A.
Partner: UNT Libraries Government Documents Department

Calculation of vacancy wind contributions in ternary diffusion

Description: In the absence of kinetic cross interactions between diffusing components, intrinsic diffusion can be described by a simple atomic mobility model. For systems where the diffusional interactions among components cannot be ignored, the interactions can be related to a vacancy wind effect in which the intrinsic flux of a component is influenced by the net vacancy flux. Atomic mobilities are calculated at selected composition points on the diffusion paths of y-phase U-Pu- Zr diffusion couples investigated at 750 {degrees}C to assess the contribution by the vacancy wind effect to the intrinsic diffusion of the individual components. The results point to the possibility that a large vacancy wind contribution may cause a component to diffuse intrinsically up its own chemical potential gradient.
Date: August 1, 1996
Creator: Petri, M.C. & Dayananda, M.A.
Partner: UNT Libraries Government Documents Department

Characterization of U-6Nb ingots produced via the electron beam cold hearth refining process

Description: A study was undertaken at Lawrence Livermore National Laboratory to characterize uranium, 6% niobium ingots produced via electron beam melting, hearth refining and continuous casting and to compare this material with conventional VIM/skull melt/VAR material. Samples of both the ingot and feed material were analyzed for niobium and trace metallic elements, carbon, oxygen and nitrogen. This material was also inspected metallographically and via microprobe analysis.
Date: November 14, 1997
Creator: McKoon, R.H.
Partner: UNT Libraries Government Documents Department

Characterization of uranium and uranium-zirconium deposits produced in electrorefining of spent nuclear fuel

Description: This paper describes the metallurgical characterization of deposits produced in molten salt electrorefining of uranium and uranium - 10.% zirconium alloy. The techniques of characterization are described with emphasis on considerations given to the radioactive and pyrophoric nature of the samples. The morphologies observed and their implications for deposit performance are also presented - samples from pure uranium deposits were comprised of chains of uranium crystals with a characteristic rhomboidal shape, while morphologies of samples from deposits containing zirconium showed more polycrystalline features. Zirconium was found to be present as a second, zirconium metal phase at or very near the uranium-zirconium dendrite surfaces. Higher collection efficiencies and total deposit weights were observed for the uranium-zirconium deposits; this performance increase is likely a result of better mechanical properties exhibited by the uranium-zirconium dendrite morphology. 18 refs., 10 figs., 1 tab.
Date: September 1, 1997
Creator: Totemeier, T.C.
Partner: UNT Libraries Government Documents Department

Uranium alloy forming process research

Description: The study of modern U-6Nb processes is motivated by the needs to reduce fabrication costs and to improve efficiency in material usage. We have studied two potential options: physical vapor deposition (PVD) for manufacturing near-net-shape U-6Nb, and kinetic-energy metallization (KEM) as a supplemental process for refurbishing recycled parts. In FY 1996, we completed two series of PVD runs and heat treatment analyses, the characterization of the microstructure and mechanical properties, a comparison of the results to data for wrought-processed material, and experimental demonstration of the KEM feasibility process with a wide range of variables (particle materials and sizes, gases and gas pressures, and substrate materials), and computer modeling calculations.
Date: March 1, 1997
Creator: Chow, T.S.; Biesiada, T.A.; Sunwoo, A.; Long, J.; Anklam, T. & Kang, S.W.
Partner: UNT Libraries Government Documents Department


Description: Several spallation experiments have been performed on the 6 wt pct alloy of uranium using gas gun driven normal plate impacts with VISAR instrumentation and soft recovery. The nominal shock pressures achieved were 28, 34, 42, 50, 55, and 82 kbar. This paper will focus on spallation modeling, e.g. using the 1 D characteristics code CHARADE to simulate the free surface particle velocity. The spallation model involves the ductile growth and coalescence of voids. Metallographical examination of recovered samples and details of the experimental apparatus are discussed in a separate paper.
Date: January 10, 2001
Creator: TONKS, D. & AL, ET
Partner: UNT Libraries Government Documents Department

Polycrystalline deformation in engineering materials: Insights from neutron diffraction during loading

Description: In-situ measurements using the non-destructive penetration of neutrons are commonplace at neutron sources and permit investigations within environmental chambers at stress, pressure, or temperature. Many of these studies explore the microstructural performance of engineering materials under service conditions. For example, by measuring phase strains during the application of static loads, neutron diffraction provides insight into failure, relaxation and load transfer mechanisms. Mechanical loading of a sample on a neutron spectrometer is usually performed with a customized load frame (small enough to fit into the typically limited available space) with the load axis horizontal. Diffraction data are recorded using detectors that surround the sample and strains are determined from changes in the measured interplanar lattice spacings in directions determined by the scattering geometry. These elastic strains indicate how the applied stress is shared throughout the microstructure. During a test, conventional strain gauges also record the macroscopic strain; that is the sum of the plastic and elastic contributions. Beyond yield the plastic contribution usually dominates the total strain but the elastic phase strains respond to the applied stress at any given load and provide clues about which phase (in a multiphase system) or which crystal orientation (in a single phase polycrystal) dictates failure.
Date: June 6, 1999
Creator: Bourke, M. & Brown, D.
Partner: UNT Libraries Government Documents Department

Development of a near-net-shape casting technology for the U-6Nb alloy. Part 1: Materials characterization, experiment design, and model construction

Description: The Oak Ridge Y-12 Plant (Y-12) is conducting highly coupled experimental and numerical studies to develop the technology needed to produce near-net-shape (NNS)-cast uranium-6 wt% niobium (U-6Nb) components which have a controlled carbon content. Current activities are focused on defining mechanical and metallurgical properties of cast material; experimental studies to define NNS casting, carbide particle flotation, and immersion-quench physics; and developing the numerical models needed to support the optimized design of NNS components. This paper summarizes the material characterization, experiment design, and model development activities.
Date: January 1, 1997
Creator: Taylor, M.J.; Keeney, J.A.; Wendel, M.W. & Demint, A.L.
Partner: UNT Libraries Government Documents Department

Irradiation testing of high density uranium alloy dispersion fuels

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

News from IPNS

Description: Niobium-base alloys are candidate materials for the dive structure in fusion reactors. For this application, the materials should exhibit resistance to aqueous corrosion, hydrogen embrittlement, and radiation damage and should have high thermal conductivity and low thermal expansion. Results of corrosion and embrittlement screening tests of several binary and ternary alloys in high-temperature water indicated that Nb-1Zr, Nb-5Mo-1Zr and Nb-5V-1Zr (wt.%) showed sufficient promise for further investigation. These alloys, together with pure Nb and Zircaloy-4, have been exposed to high-purity water containing a low concentration of dissolved oxygen (<12 ppb) at 170, 230, and 300C for up to {approx}3200 h. Weight-change data, microstructural observations, and qualitative mechanical-property evaluations reveal that Nb-5V-1Zr is the most promising alloy at higher temperatures. Below {approx}200C the alloys exhibit similar corrosion behavior.
Date: December 31, 1995
Creator: Brown, B.S.
Partner: UNT Libraries Government Documents Department

Experimental and calculated swelling behavior of U-10 wt.% Mo under low irradiation temperatures.

Description: SEM micrographs of U-10 wt.% Mo irradiated at low temperature in the ATR to about 40 at. % burnup show the presence of cavities. We have used a rate-theory-based model to investigate the nucleation and growth of cavities during low-temperature irradiation of uranium-molybdenum alloys in the presence of irradiation-induced interstitial-loop formation and growth. In addition, the evolution of forest dislocations was calculated based on dislocation loop growth and simultaneous climb and glide of unfaded loops. Consolidation of the dislocation structure takes into account capture of interstitial dislocation loops and annihilation of adjacent dislocations, as well as loss to grain boundaries. A di-interstitial is assumed to be the nucleus of a dislocation loop. Cavities are nucleated when two gas atoms come together in the presence of at least one vacancy. Cavity growth occurs by the influx of gas atoms and/or vacancies. In turn, the free interstitial concentration, and thus (due to recombination) the free-vacancy concentration, depends on the dislocation density. Bias-driven growth of cavities can lead to substantial swelling of the alloy (void swelling). However, our calculations indicate that the swelling mechanism in the U-10 wt.% Mo alloy at low irradiation temperatures is fission gas driven. The calculations also indicate that the observed bubbles must be associated with a sub-grain structure. Calculated swelling and bubble-size-distribution are compared with irradiation data.
Date: September 29, 1998
Creator: Rest, J.
Partner: UNT Libraries Government Documents Department

Modeling of material and energy flow in an EBCHR casting system

Description: A numerical and experimental analysis is made of fluid flow and heat transfer in a continuous casting system with an electron-beam energy source. For a cylindrical ingot confined in a water-cooled crucible, a two-dimensional, steady-state model is developed which includes the effects of free convection in the pool and conduction in the two-phase and solid regions. A modified Galerkin finite element method is used to solve for the flow and temperature fields simultaneously with the upper and lower boundaries of the pool. The calculation grid deforms along vertical spines as these phase boundaries move. Heat flows are measured in a steady-state experiment involving a short ingot and no pouring. Heat transfer coefficients representing contact resistance are determined, and measured heat flows are compared with model values. Flow and temperature fields along with solidification-zone boundaries are calculated for the experimental case and a case in which the ingot cooling is improved.
Date: November 1, 1994
Creator: Westerberg, K.W. & McClelland, M.A.
Partner: UNT Libraries Government Documents Department

Thermal simulation of quenching uranium-0. 75% titanium alloy in water

Description: A computer model, The Quench Simulator, has been developed to simulate and predict in detail the behavior of U-0.75 Ti alloy when quenched at high temperature (about 850/sup 0/C) in cold water. The code allows one to determine the time- and space-dependent distributions of temperature, residual stress, distortion, and microstructure that evolve during the quenching process. The nonlinear temperature- and microstructure-dependent properties, as well as the cooling rate-dependent heats of transformation, are incorporated into the model. The complex boiling heat transfer with its various regimes and other thermal boundary conditions are simulated. Experiments have been performed and incorporated into the model. Both sudden submersion and gradual controlled immersion can be applied. A parametric and sensitivity study has been performed demonstrating the importance of the thermal boundary conditions applied for achieving certain product characteristics. The thermal aspects of the model and its applications are discussed and demonstrated.
Date: January 1, 1985
Creator: Siman-Tov, M.; Llewellyn, G.H.; Childs, K.W.; Ludtka, G.M. & Aramayo, G.A.
Partner: UNT Libraries Government Documents Department