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Equilibrium Dissociation Pressures of the Delta and Epsilon Phases in the Zirconium-Hydrogen System

Description: Abstract: Pressure-temperature isochores were obtained for zirconium-hydrogen alloys, spanning the H/Zr composition range of 1.430 to 1.910. The studies were confined to the temperature limits of 300 to 900ÂșC, and the pressure limits of 0.01 to 10.0 atm.
Date: May 31, 1964
Creator: Raymond, J. W.
Partner: UNT Libraries Government Documents Department

Deformation Microstructures and Creep Mechanisms in Advanced ZR-Based Cladding Under Biazal Loading

Description: Investigate creep behavior of Zr-based cladding tubes with attention to basic creep mechanisms and transitions in them at low stresses and/or temperatures and study the dislocation microstructures of deformed samples for correlation with the underlying micromechanism of creep
Date: August 11, 2008
Creator: Murty, K. Linga (KL)
Partner: UNT Libraries Government Documents Department

Study of Second Phase Particles and Fe content in Zr Alloys Using the Advanced Photon Source at Argonne

Description: We have conducted a study of second phase particles and matrix alloying element concentrations in zirconium alloys using synchrotron radiation from the Advanced Photon Source (APS) at Argonne National Laboratory. The high flux of synchrotron radiation delivered at the 2BM beamline compared to conventional x-ray generators, enables the detection of very small precipitate volume fractions. We detected the standard C14 hcp Zr(Cr,Fe)2 precipitates, (the stable second phase in Zircaloy-4) in the bulk material at a cumulative annealing parameter as low as 10-20 h, and we followed the kinetics of precipitation and growth as a function of the cumulative annealing parameter (CAP) in the range 10-22 (quench) to 10-16 h. In addition, the unique combination of spatial resolution and elemental sensitivity of the 2ID-D/E microbeam line at the Advanced Photon Source at Argonne (APS) allows study of the alloying element concentrations at ppm levels in an area as small as 0.2 mm. We used x-ray fluorescence induced by this sub-micron x-ray beam to determine the concentration of these alloying elements in the matrix as a function of alloy type and thermal history. We discuss these results and the potential of synchrotron radiation-based techniques for studying zirconium alloys.
Date: November 7, 2001
Creator: Motta, Arthur T.
Partner: UNT Libraries Government Documents Department

A REPORT ON SOME ATTEMPTS TO CAST CENTRIFUGALLY FUEL ELEMENTS OF SMALL DIAMETER

Description: The applicability of the centrifugal casting technique to the production of multiple castings of fuel pins of small diameter and of thin fuel plates was investigated. Fuel pins measuring 0.185 in. in diameter by 4 1/4 in. long of unalloyed uranium and of a uranium-2 wt.% zirconium alloy were cast successfully in batches of sixteen pins per melt. Sixteen different metals and alloys were used as mold materials. Smaller and longer fuel pins, 0.165 in. in diameter by 9 3/4 in. long, of similar compositions were cast successfully in brass and copper molds. Thirty-six pins of the same diameter and length were cast simultaneously in each casting run. Attempts to cast centrifugally thin uranium plates measuring 9 in. long by 2 in. wide by 0.04 in. thick proved to be only partially successful, but encouraging. These plates were cast into graphite molds at the rate of six plates per run. The maximum usable length of the unalloyed uranium plates cast did not exceed six inches. (auth)
Date: March 1, 1961
Creator: Yaggee, F.L.
Partner: UNT Libraries Government Documents Department

RECOIL RANGE OF FISSION FRAGMENTS IN ZIRCONIUM

Description: The recoil ranges of fission fragruents in zirconium were measured by irradiating small pieces of a Zircaloy-2 ribbon containing 492 ppm of homogeneously dispersed U/sup 235/. The samples were wrapped in aluminum foil during the irradiation to catch the fission fragments escaping from the surface of the Zircaloy-2. Average values of 10.1 and 8.15 microns were obtained for the recoil ranges of the median light and heavy fission fragments, respectively; these values compare quite well with values calculated from empirical equations. (auth)
Date: November 1, 1959
Creator: Smith, E.R. & Frank, P.W.
Partner: UNT Libraries Government Documents Department

Steady State Creep of Zirconium at High and Intermediate Temperatures

Description: Creep of zirconium and zirconium alloys has been labeled ''anomalous.'' Researchers often report that zirconium and its alloys never reach true steady state creep and have stress exponents that continuously change with stress and temperature. Many varied interpretations have been offered explaining the creep behavior of zirconium. Some have suggested that creep is diffusion controlled, while others maintain that creep is dislocation glide controlled. Cumulative zirconium creep data will be presented based on an extensive literature review. An interpretation of results will be presented and compared to previous interpretations.
Date: April 8, 2000
Creator: Rosen, R.S. & Hayes, T.A.
Partner: UNT Libraries Government Documents Department

Microhardness and elastic modulus of nanocrystalline Al-Zr

Description: An investigation of the mechanical properties of nanocrystalline Al-Zr alloy composites has been conducted via nanoindentation and Vickers microhardness experiments. The microhardness of the samples exhibits a four-fold increase over the concentration range of 0-30 wt.% Zr, from {approximately}0.7 GPa to nearly 3 GPa. The aluminum grain size is found to be strongly correlated with the level of zirconium present in the samples, suggesting that the observed hardness increase can be attributed to the combined effects of alloying and grain size reduction. The elastic moduli of the nanocrystalline Al-Zr samples are determined to be similar to the modulus of coarse-grained aluminum and independent of zirconium content.
Date: November 1, 1995
Creator: Rittner, M.N.; Weertman, J.R. & Eastman, J.A.
Partner: UNT Libraries Government Documents Department

Mechanical properties of nanocrystalline metals, intermetalics and multiphase materials determined by tension, compression and disk-bend techniques

Description: The mechanical behavior of nanocrystalline metallic, intermetallic, and multiphase materials was investigated using tension, compression, and disk-bend techniques. Nanocrystalline NiAl, Al-Al{sub 3}Zr, and Cu were synthesized by gas condensation and either resistive or electron beam heating followed by high temperature vacuum compaction. Disk- bend tests of nanocrystalline NiAl show evidence of improved ductility at room temperature in this normally extremely brittle material. In contrast, tension tests of multiphase nanocrystalline Al- Al{sub 3}Zr samples show significant increases in strength by substantial reductions in ductility with decreasing grain size. Compression tests of nanocrystalline copper result in substantially higher yield stress and total elongation values than those measured in tensile tests. Implications for operative deformation mechanisms in these materials are discussed.
Date: February 1, 1997
Creator: Eastman, J.A.; Thompson, L.J.; DiMelfi, R.J.; Choudry, M.; Dollar, M.; Weertman, J.R. et al.
Partner: UNT Libraries Government Documents Department

Fundamental studies of ceramic/metal interfacial reactions at elevated temperatures.

Description: This work characterizes the interfaces resulting from exposing oxide and non-oxide ceramic substrates to zirconium metal and stainless steel-zirconium containing alloys. The ceramic/metal systems together were preheated at about 600 C and then the temperatures were increased to the test maximum temperature, which exceeded 1800 C, in an atmosphere of high purity argon. Metal samples were placed onto ceramic substrates, and the system was heated to elevated temperatures past the melting point of the metallic specimen. After a short stay at the peak temperature, the system was cooled to room temperature and examined. The chemical changes across the interface and other microstructural developments were analyzed with energy dispersive spectroscopy (EDS). This paper reports on the condition of the interfaces in the different systems studied and describes possible mechanisms influencing the microstructure.
Date: December 14, 2000
Creator: McDeavitt, S. M.; Billings, G. W. & Indacochea, J. E.
Partner: UNT Libraries Government Documents Department

A Report on Some Attempts to Cast Centrifugally Fuel Elements of Small Diameter

Description: Report issued by the Argonne national Laboratory discussing centrifugal casting of thin uranium plates. The applicability of the centrifugal casting technique to the production of multiple castings of fuel pins of small diameter and of thin fuel plates was investigated. This report includes tables, illustrations, and photographs.
Date: March 1961
Creator: Yaggee, F. L.
Partner: UNT Libraries Government Documents Department

U-Pu-Zr Metal Alloy: a Potential Fuel for LMFBR's

Description: This report critically reviews the available information pertinent to the potential use of uranium-plutonium-zirconium alloy fuels clad with stainless steel in LMFBR's. The areas considered include breeding potential, burnup potential, thermal performance, fuel fabricability, fuel reprocessing, and safety considerations. Because information on uranium-plutonium-zirconium alloys is limited, wide use is made of experience with EBR-II metallic driver fuel to infer advantages and limitations of uranium-plutonium-zirconium fuels. It is concluded that sufficient potential exists for the applicability of uranium-plutonium-zirconium fuels to LMBFR's to warrant additional analytical and experimental studies.
Date: November 1975
Creator: Walter, C. M.; Golden, G. H. & Olson, N. J.
Partner: UNT Libraries Government Documents Department

Uniaxial Tensile Properties of Zircaloy Containing Oxygen : Summary Report

Description: The uniaxial stress-strain behavior of Zircaloy-2 and -4, Zircaloy-oxygen alloys with a uniform oxygen distribution, and composite specimens with a ZrO2/alpha/beta layer structure was investigated over the range of experimental conditions: temperature 25-1400 degrees C; strain rate ; oxygen content 0.11 - 4.4 wt %; grain size 5-50 micrometers; texture longitudinal, transverse, and diagonal orientations; and microstructural state, which consists of the equiaxed alpha phase and various transformed beta acicular structures. The work-hardening and strain-rate sensitivity parameters were determined from the experimental results, and the tensile properties were correlated with oxygen concentration, oxygen distribution in the material, and microstructure. Dynamic strain-aging phenomena were observed in Zircaloy at 200, 400, and 700 degrees C, and super-plastic deformation occurred at 850 and 1000 degrees C. An increase in the oxygen concentration in homogeneous Zircaloy-oxygen alloys increased the ultimate tensile strength and decreased the total strain, particularly below approximately 900/sup 0/C. In composite specimens with the ZrO2/alpha/beta structure, the total oxygen content had little effect on the ultimate tensile strength below approximately 1000 degrees C, but the strength increased with oxygen content at higher temperatures. Information on the effects of grain size, oxygen content, texture, and strain rate on the stress-strain behavior suggests that the dominant mechanism of super-plastic deformation in Zircaloy near approximately 850/sup 0/C is grain-boundary sliding at the alpha-beta interface with accommodation by diffusional creep, dislocation slip, and grain-boundary migration. Good correlation was obtained between ductility and values of the strain-rate sensitivity parameter.
Date: June 1977
Creator: Garde, A. M.; Chung, H. M. & Kassner, T. F.
Partner: UNT Libraries Government Documents Department

Fuel Cycle Programs, Quarterly Progress Report: April-June 1981

Description: Quarterly report of the Argonne National Laboratory Chemical Engineering Division regarding activities related to properties and handling of radioactive materials, operation of nuclear reactors, and other relevant research.
Date: March 1982
Creator: Steindler, M. J.; Vogler, Seymour; Vandegrift, G. F.; Williams, Jacqueline; Gerding, T. J.; Jardine, L. J. et al.
Partner: UNT Libraries Government Documents Department

Management of Waste Cladding Hulls

Description: This report reviews experience and research related to the pyrophoricity of zirconium and zirconium alloys. The results of recent investigations of the behavior of Zircaloy and some observations of industrial handling and treatment of Zircaloy tubing and scrap are also discussed. A model for the management of waste Zircaloy cladding hulls from light water reactor fuel reprocessing is offered, based on an evaluation of the reviewed information. It is concluded that waste Zircaloy cladding hulls do not constitute a pyrophoric hazard if, following the model flow sheet, finely divided metal is oxidized during the management procedure. Steps alternative to the model are described which yield zirconium in deactivated form and also accomplish varying degrees of transuranic decontamination. Information collected into appendixes is (1) a collation of zirconium pyrophoricity data from the literature, (2) calculated radioactivity contents in Zircaloy cladding hulls from spent LWR fuels, and (3) results of a laboratory study on volatilization of zirconium from Zircaloy using HCl or chlorine.
Date: November 1977
Creator: Kullen, B.; Levitz, N. M. & Steindler, M. J.
Partner: UNT Libraries Government Documents Department

Applied Physical Chemistry Progress Report, October 1991 - September 1992

Description: This document reports on the work done in applied physical chemistry at the Chemical Technology Division (CMT), Argonne National Laboratory (ANL), in the period October 1991 through September 1992. this work includes research into the process that control the release and transport of fission products under accident-like conditions in a light water reactor, the thermophysical properties of the metal fuel in the Integral Fast Reactor under development at ANL, and the properties of candidate tritium breeding materials in environments simulating those of fusion energy systems.
Date: December 1993
Creator: Johnson, C. E.
Partner: UNT Libraries Government Documents Department

Review of the oxidation rate of zirconium alloys.

Description: The oxidation of zirconium alloys is one of the most studied processes in the nuclear industry. The purpose of this report is to provide in a concise form a review of the oxidation process of zirconium alloys in the moderate temperature regime. In the initial ''pre-transition'' phase, the surface oxide is dense and protective. After the oxide layer has grown to a thickness of 2 to 3 {micro}m's, the oxidation process enters the ''post-transition'' phase where the density of the layer decreases and becomes less protective. A compilation of relevant data suggests a single expression can be used to describe the post-transition oxidation rate of most zirconium alloys during exposure to oxygen, air, or water vapor. That expression is: Oxidation Rate = 13.9 g/(cm{sup 2}-s-atm{sup -1/6}) exp(-1.47 eV/kT) x P{sup 1/6} (atm{sup 1/6}).
Date: November 1, 2005
Creator: Causey, Rion A. (Sandia National Laboratories, Livermore, CA); Cowgill, Donald F. (Sandia National Laboratories, Livermore, CA) & Nilson, Robert H. (Sandia National Laboratories, Livermore, CA)
Partner: UNT Libraries Government Documents Department

Underground Corrosion of Activated Metals, 6-Year Exposure Analysis

Description: The subsurface radioactive disposal site located at the Idaho National Laboratory contains neutronactivated metals from non-fuel nuclear-reactor-core components. A long-term underground corrosion test is being conducted to obtain site-specific corrosion rates to support efforts to more accurately estimate the transfer of activated elements in the surrounding arid vadose zone environment. The test uses nonradioactive metal coupons representing the prominent neutron-activated materials buried at the disposal location, namely, Type 304L stainless steel (UNS S30403), Type 316L stainless steel (S31603), nickel-chromium alloy (UNS NO7718), beryllium, aluminum 6061-T6 (A96061), and a zirconium alloy (UNS R60804). In addition, carbon steel (the material presently used in the cask disposal liners and other disposal containers) and a duplex stainless steel (UNS S32550) are also included in the test. This paper briefly describes the ongoing test and presents the results of corrosion analysis from coupons exposed underground for 1, 3, and 6 years.
Date: March 1, 2006
Creator: Flitton, M. K. Adler & Yoder, T. S.
Partner: UNT Libraries Government Documents Department