DOCSLIB.ORG

Anisotropic Thermal Conductivity in Uranium Dioxide

Total Page：16

File Type：pdf, Size：1020Kb

Download full-text PDF Read full-text
Anisotropic Thermal Conductivity in Uranium Dioxide ARTICLE Received 9 May 2014 | Accepted 27 Jun 2014 | Published 1 Aug 2014 DOI: 10.1038/ncomms5551 Anisotropic thermal conductivity in uranium dioxide K. Gofryk1,w,S.Du2,w, C.R. Stanek3, J.C. Lashley1, X.-Y. Liu3, R.K. Schulze1, J.L. Smith1, D.J. Safarik1, D.D. Byler3, K.J. McClellan3, B.P. Uberuaga3, B.L. Scott4 & D.A. Andersson3 The thermal conductivity of uranium dioxide has been studied for over half a century, as uranium dioxide is the fuel used in a majority of operating nuclear reactors and thermal conductivity controls the conversion of heat produced by fission events to electricity. Because uranium dioxide is a cubic compound and thermal conductivity is a second-rank tensor, it has always been assumed to be isotropic. We report thermal conductivity measurements on oriented uranium dioxide single crystals that show anisotropy from 4 K to above 300 K. Our results indicate that phonon-spin scattering is important for understanding the general thermal conductivity behaviour, and also explains the anisotropy by coupling to the applied temperature gradient and breaking cubic symmetry. 1 Materials Technology-Metallurgy, Materials Science and Technology Division, Los Alamos National Laboratory, Los Alamos, New Mexico 87545, USA. 2 Physics and Chemistry of Materials, Theoretical Division, Los Alamos National Laboratory, Los Alamos, New Mexico 87545, USA. 3 Materials Science in Radiation & Dynamical Extremes, Materials Science and Technology Division, Los Alamos National Laboratory, Los Alamos, New Mexico 87545, USA. 4 Materials Physics and Applications Division, Los Alamos National Laboratory, Los Alamos, New Mexico 87545, USA. w Present address: Fuel Modeling and Simulation Department, Idaho National Laboratory, P.O. Box 1625, Idaho Falls, Idaho 83415, USA (K.G.); Present address: Division of Functional Materials and Nanodevices, Ningbo Institute of Materials Technology and Engineering, Chinese Academy of Sciences, Ningbo, Zhejiang 315201, China (S.D.). Correspondence and requests for materials should be addressed to C.R.S. (email: [email protected]) or to D.A.A. (email: [email protected]). NATURE COMMUNICATIONS | 5:4551 | DOI: 10.1038/ncomms5551 | www.nature.com/naturecommunications 1 & 2014 Macmillan Publishers Limited. All rights reserved. ARTICLE NATURE COMMUNICATIONS | DOI: 10.1038/ncomms5551 oday nearly 20% of the world’s electricity is generated by originated from molecular dynamics (MD) simulations using 6–8 nuclear energy from UO2 fuel. The thermal conductivity of the direct method , which calculates thermal conductivity by TUO2 governs the conversion of heat produced from fission measuring the temperature gradient in a rectangular cuboid events into electricity and it is an important parameter in reactor with heat added and removed at a hot and cold plate, respectively. design and safety. Scatter among thermal conductivity (k) data The simulations predict large differences for the effective from previous experimental studies on UO2 single crystals cannot thermal conductivity in the o1004 o1104 and o1114 easily be explained by sample impurities or inherent measure- crystallographic directions at 300 K, see Fig. 1a. In an attempt to ment errors. (see ref. 1 and references therein). We hypothesize eliminate boundary condition artifacts related to scattering by the that much of this scatter is due to hidden anisotropy, overlooked hot and cold plates, the thermal conductivity is obtained for in previous experiments that did not specify single crystal infinite simulationÀÁ cell length by extrapolation7,8. Figure 1a shows À 1 À 1 orientation. The crystal symmetry of materials dictates the that k ¼ fLz (Lz is the length of the simulation cell) does symmetry of the response to applied potential gradients. not obey the expected linear relationship for UO2 in the o1004 Thermal conductivity is a second-rank tensor (that is, k ¼ kij) direction, and is instead quadratic. The extrapolated values at that governs the flow of energy under an applied temperature 300 K in the inset of Fig. 1a (based on the quadratic fits) show gradient. For materials with cubic point and space group symmetry that the anisotropy between o1114 and o1004 is B15%. This it is isotropic. Thermal conductivity is almost exclusively result is noteworthy as the quadratic term indicates that the dominated by phonons in insulators and semiconductors. At intrinsic material properties are affected, whereas a linear term very low temperatures, where ballistic phonon transport may would correspond to a mere change in the phonon scattering occur, exceptions to the isotropy of cubic materials have been length. The underlying physics is illustrated by the projected MD reported, for example, ref. 2. This can be explained by the thermal trajectory of an oxygen ion, see Fig. 1b. This simulation was conductivity being a fourth rank tensor (k ¼ kijlm) related to the performed in the same rectangular cuboid cell as the thermal elastic stiffness tensor when phonon scattering is minimal3. conductivity simulations, but without a temperature gradient. Here we show by experiments on single crystals that Although the average position coincides with the cubic UO2 UO2, which crystallizes in the cubic fluorite structure ðÞFm3m , fluorite structure, the shape of the distribution is an ellipsoid exhibits anisotropic thermal conductivity at temperatures at rather than the expected sphere. The ellipsoidal character is still which ballistic mechanisms are not possible. The anisotropy is present during short time intervals in a cubic simulation cell, rather due to the applied temperature gradient interacting with however it averages to a sphere since the ellipsoid orientation the electric moments on uranium ions thus affecting the phonon- along the crystallographic axes is uncorrelated in space and time. spin interactions to slightly break cubic symmetry. This effect is Apparently, the trajectory is sensitive to small perturbations, such caused by the magnetic properties of UO2 and the degenerate as the dimensionality of the simulation cell or the applied bands giving rise to Jahn–Teller distortions. This result is temperature gradient. These perturbations break the underlying unexpected since the crystal structure for both the paramagnetic cubic symmetry, and lead to a preferential ellipsoid orientation and the antiferromagnetic phases of UO2 are cubic and there is along one of the three o1004 axes. This orientation bias gives no clear evidence of non-cubic distortions in diffraction rise to anisotropy in the physical properties. experiments4. These results should also apply to the inter- pretation of other magnetic strongly correlated materials 5 exhibiting Jahn–Teller physics . Measurements of anisotropic UO2 thermal conductivity. These MD results inspired a detailed experimental study of the UO2 Results thermal conductivity as a function of temperature with a thermal Molecular dynamics simulations of UO2 thermal conductivity. gradient applied in different crystallographic directions on Our consideration of anisotropic UO2 thermal conductivity oriented single crystal bars. We find that the overall shape of the ab61.8 –1 –1 ∞=20.6 Wm K 0.10 1.00 〈111〉 Exp., Calc. 0.95 〈110〉 Exp., Calc. 61.6 300 K 〈100〉 Exp., Calc. –1 ) –1 0.08 0.90 61.4 0.85 (mKW 300 [001] (Å) –1 0.06 T (K) –1 –1 ∞=24 Wm K 61.2 〈 〉 0.04 100 〈111〉 61.0 0.00 0.01 0.02 0.03 0.04 0.05 0.06 0.07 1.0 1.2 1.4 1.6 1.8 –1 –1 [100] (Å) Lz (nm ) ÀÁ À 1 À 1 Figure 1 | Anisotropic UO2 thermal conductivity from MD simulations. (a) The thermal conductivity (k) variation (k ) as a function of the length Lz of the simulation cell for the temperature gradient applied in the o1004 and o1114 lattice directions at 300 K. The lines represent linear and quadratic 7,8 fits to the data used for extrapolation to infinite length (kN) of the simulation cells . The inset compares the experimentally measured thermal conductivity values (Exp.) in the o1004 o1104 and o1114 crystallographic directions with the MD simulations (Calc.). The length of the MD simulation cells were B30 nm for these data points. The values have been normalized to the highest thermal conductivity value at 300 K (k300). (b) Projected trajectory of one oxygen ion during a 500 ps MD run in the rectangular cuboid simulation cell at 300 K. The MD simulation was performed in the NVE ensemble without an applied thermal gradient. The horizontal axis is the [100] and the vertical axis the [001] direction of the fluorite lattice. The elongated direction of the simulation cell is along the vertical axis. The ellipsoid character of the distribution is emphasized by the dashed red line as compared with the white line representing the spherical distribution. 2 NATURE COMMUNICATIONS | 5:4551 | DOI: 10.1038/ncomms5551 | www.nature.com/naturecommunications & 2014 Macmillan Publishers Limited. All rights reserved. NATURE COMMUNICATIONS | DOI: 10.1038/ncomms5551 ARTICLE thermal conductivity is atypical for insulating crystals without same energy range as the occupied phonons, otherwise scattering defects, see Fig. 2. For all directions, pronounced double-peak is negligible10,16–18. There are no excitation levels in the triplet behaviour in the thermal conductivity is observed with maxima ground state of uranium ions above TN that can interact with occurring at B10 and 220 K and a minimum at the Ne´el tem- phonons19,20. However, inelastic neutron scattering experiments 1,9 perature TN ¼ 30.8 K, as reported previously . This behaviour have identified a dynamic Jahn–Teller distortion in o1004 has been observed in other materials with antiferromagnetic lattice directions that splits the ground state into three singlets transitions9–13, which is due to resonant scattering of phonons with dispersive peaks of magnetic character between 3 and through excitations of the spins on the magnetic ions in the 10 meV19,20. These excitations are in the appropriate energy paramagnetic phase and absence of this scattering mechanism range for interaction with phonons and give rise to significant 9–13 below TN due to spin ordering .
Load more

Recommended publications
  • Measurement of Uranium Dioxide Thermophysical Properties by the Laser Flash Method
    2009 International Nuclear Atlantic Conference - INAC 2009 Rio de Janeiro,RJ, Brazil, September27 to October 2, 2009 ASSOCIAÇÃO BRASILEIRA DE ENERGIA NUCLEAR - ABEN ISBN: 978-85-99141-03-8 MEASUREMENT OF URANIUM DIOXIDE THERMOPHYSICAL PROPERTIES BY THE LASER FLASH METHOD Pablo Andrade Grossi 1, Ricardo Alberto Neto Ferreira 2, Denise das Mercês Camarano 3, Roberto Márcio de Andrade 4 1,2,3 Centro de Desenvolvimento da Tecnologia Nuclear-CDTN Comissão Nacional de Energia Nuclear-CNEN Av. Presidente Antônio Carlos 6627 Cidade Universitária-Pampulha-Caixa Postal 941 CEP: 31.161-970 Belo Horizonte, MG 1 [email protected] 2 [email protected] 3 [email protected] 4 Departamento de Engenharia Mecânica Escola de Engenharia da Universidade Federal de Minas Gerais Av. Presidente Antônio Carlos 6627-Cidade Universitária-Pampulha CEP: 31.270-901 - Belo Horizonte – MG - Brasil 4 [email protected] ABSTRACT The evaluation of the thermophysical properties of uranium dioxide (UO 2), including a reliable uncertainty assessment, are required by the nuclear reactor design. These important informations are used by thermohydraulic codes to define operational aspects and to assure the safety, when analyzing various potential situations of accident. The Laser Flash Method had become the most popular method to measure the thermophysical properties of materials. Despite its several advantages, some experimental obstacles have been found due to the difficulty to obtain experimentally the ideals initial and boundary conditions required by the original method. An experimental apparatus and a methodology for estimating uncertainties of thermal diffusivity, thermal conductivity and specific heat measurements based on the Laser Flash Method are presented. A stochastic thermal diffusion modeling has been developed and validated by Standard Samples.
    [Show full text]
  • Depleted Uranium Technical Brief
    Disclaimer - For assistance accessing this document or additional information,please contact [email protected]. Depleted Uranium Technical Brief United States Office of Air and Radiation EPA-402-R-06-011 Environmental Protection Agency Washington, DC 20460 December 2006 Depleted Uranium Technical Brief EPA 402-R-06-011 December 2006 Project Officer Brian Littleton U.S. Environmental Protection Agency Office of Radiation and Indoor Air Radiation Protection Division ii iii FOREWARD The Depleted Uranium Technical Brief is designed to convey available information and knowledge about depleted uranium to EPA Remedial Project Managers, On-Scene Coordinators, contractors, and other Agency managers involved with the remediation of sites contaminated with this material. It addresses relative questions regarding the chemical and radiological health concerns involved with depleted uranium in the environment. This technical brief was developed to address the common misconception that depleted uranium represents only a radiological health hazard. It provides accepted data and references to additional sources for both the radiological and chemical characteristics, health risk as well as references for both the monitoring and measurement and applicable treatment techniques for depleted uranium. Please Note: This document has been changed from the original publication dated December 2006. This version corrects references in Appendix 1 that improperly identified the content of Appendix 3 and Appendix 4. The document also clarifies the content of Appendix 4. iv Acknowledgments This technical bulletin is based, in part, on an engineering bulletin that was prepared by the U.S. Environmental Protection Agency, Office of Radiation and Indoor Air (ORIA), with the assistance of Trinity Engineering Associates, Inc.
    [Show full text]
  • High-Density Concrete with Ceramic Aggregate Based on Depleted Uranium Dioxide
    HIGH-DENSITY CONCRETE WITH CERAMIC AGGREGATE BASED ON DEPLETED URANIUM DIOXIDE S.G. Ermichev, V.I. Shapovalov, N.V.Sviridov (RFNC-VNIIEF, Sarov, Russia) V.K. Orlov, V.M. Sergeev, A. G. Semyenov, A.M. Visik, A.A. Maslov, A. V. Demin, D.D. Petrov, V.V. Noskov, V. I. Sorokin, O. I. Uferov (VNIINM, Moscow, Russia) L. Dole (ORNL, Oak Ridge, USA) Abstract - Russia is researching the production and testing of concretes with ceramic aggregate based on depleted uranium dioxide (UO2). These DU concretes are to be used as structural and radiation-shielded material for casks for A-plant spent nuclear fuel transportation and storage. This paper presents the results of studies aimed at selection of ceramics and concrete composition, justification of their production technology, investigation of mechanical properties, and chemical stability. This Project is being carried out at the A.A. Bochvar All-Russian Scientific-Research Institute of inorganic materials (VNIINM, Russia) and Russian Federal Nuclear Center – All-Russia Scientific Research Institute of Experimental Physics (RFNC-VNIIEF, Russia). This Project #2691 is financed by the International Science and Technology Center (ISTC) under collaboration of the Oak- Ridge National Laboratory (USA) I. INTRODUCTION The current practice of ensuring the required gamma conduction, thermo-, radiation and corrosion resistance, shielding and strength of metal-concrete casks is based on service life and water resistance. concrete density. For this purpose high-density rocks Specific characteristics of UO2’s chemical activity (magnetite, iron glance, barium sulfate, etc.) as well as because of its small size of particles and thus great specific scrap, scale, broken metal chips and others are introduced surface area prevent the use of traditional methods of UO2 into concrete as coarse aggregates.
    [Show full text]
  • I -F??TI O08'w –Gçanooog –Ggan
    Feb. 27, 1962 L. BURRIS, JR., ETA 3,023,097 REPROCESSING URANIUM DIOXDE FUES Filed Nov. 23, 1959 2 Sheets-Sheet I-F??TI OOOG–ggan O08’w–gçan no?opa8 (g?20%OG) INVENTOR. Les lie 8urris, Jr. Af r e di S c h n e i der Attorney Feb. 27, 1962 L. BURRIS, JR., ETA 3,023,097 REPROCESSING URANIUM DIOXDE FUELS Filed Nov. 23, 1959 2 Sheets-Sheet 2 - 4O - 5 ? - 6 O - 8O - 9 O - OO - , ? - 2 O - 3 O Nd 2O Gnd Sm2O3 - 4 O - 150 Pr2O3 and Ce2O3 O 3OO 5OO IOOO 5OO 2OOO 25 OO Temperature (K) F I Leslie Burris,INVENTOR. Jr. Alfred Schneid er BY ????? C???-e Attorney - 3,023,097 United States Patent Office Patented Feb. 27, 1962 2 of the dioxide fuel and its commingled substances in prep 3,323,097 aration for the next step about to be described; this size REPROCESSING GUARNUM DIXE FUELS Lesie Burris, Jr., and Alfred Schneider, Naperville, H., reduction is best carried out by successive oxidations and aSsigi nors to the United States of America as rere reductions which cause crumbling by reason of the suc sented by the United States Atomic Energy Cons cessive changes in crystal structure. After the particle ?$$??? size has been reduced sufficiently, a molten reducing i Fied Nov. 23, 1959, Ser. No. 854,989 metal of the group consisting of zinc and cadmium is 4 Claims. (C. 75-84.) introduced to the crumbled mixture, and this causes quite a number of metals to go into the metallic state, where The invention relates to a novel pyrometalurgical O upon they alloy with the reducing metal and the molten method of reprocessing uranium dioxide fuels after they metal phase may then be separated from the solid oxide have become contaminated by ª fission products in i nu phase consisting of the oxides of metals such as uranium, clear reactors, and more particularly, to a novel method rare earths and plutonium which are still not reduced.
    [Show full text]
  • Dissolution of Uranium Dioxide in Nitric Acid Media: What Do We Know?
    EPJ Nuclear Sci. Technol. 3, 13 (2017) Nuclear © Sciences P. Marc et al., published by EDP Sciences, 2017 & Technologies DOI: 10.1051/epjn/2017005 Available online at: http://www.epj-n.org REGULAR ARTICLE Dissolution of uranium dioxide in nitric acid media: what do we know? Philippe Marc1, Alastair Magnaldo1,*, Aimé Vaudano1, Thibaud Delahaye2, and Éric Schaer3 1 CEA, Nuclear Energy Division, Research Department of Mining and Fuel Recycling Processes, Service of Dissolution and Separation Processes, Laboratory of Dissolution Studies, 30207 Bagnols-sur-Cèze, France 2 CEA, Nuclear Energy Division, Research Department of Mining and Fuel Recycling Processes, Service of Actinides Materials Fabrication, Laboratory of Actinide Conversion Processes, 30207 Bagnols-sur-Cèze, France 3 Laboratoire Réactions et Génie des Procédés, UMR CNRS 7274, University of Lorraine, 54001 Nancy, France Received: 16 March 2016 / Received in final form: 15 November 2016 / Accepted: 14 February 2017 Abstract. This article draws a state of knowledge of the dissolution of uranium dioxide in nitric acid media. The chemistry of the reaction is first investigated, and two reactions appear as most suitable to describe the mechanism, leading to the formation of monoxide and dioxide nitrogen as reaction by-products, while the oxidation mechanism is shown to happen before solubilization. The solid aspect of the reaction is also investigated: manufacturing conditions have an impact on dissolution kinetics, and the non-uniform attack at the surface of the solid results in the appearing of pits and cracks. Last, the existence of an autocatalytic mechanism is questionned. The second part of this article presents a compilation of the impacts of several physico-chemical parameters on the dissolution rates.
    [Show full text]
  • Experimental Studies of Neutron Irradiated Uranium Dioxide at High Temperatures
    NL90C1051 lNIS-mi— 12739 Experimental studies of neutron irradiated uranium dioxide at high temperatures R.H.J. Tanke EXPERIMENTAL STUDIES OF NEUTRON IRRADIATED URANIUM DIOXIDE AT HIGH TEMPERATURES EXPERIMENTELE STUDIES VAN MET NEUTRONEN BESTRAALD URANIUMDIOXIDE BIJ HOGE TEMPERATUREN (met een samenvatting in het Nederlands) PROEFSCHRIFT TER VERKRIJGING VAN DE GRAAD VAN DOCTOR AAN DE RIJKSUNIVERSITEIT TE UTRECHT, OP GEZAG VAN DE RECTOR MAGNIFICUS PROF.DR. J.A. VAN GÏNKEL, VOLGENS BESLUIT VAN HET COLLEGE VAN DEKANEN IN HET OPENBAAR TE VERDEDIGEN OP WOENSDAG 20 JUNI 1990 DES NAMIDDAGS TE 4.15 UUR. door Richardus Hendrikus Johannes Tanke geboren op 19 april 1957 te Hengelo (O) Promotor: Prof.Dr. CD. Andriesse Aan Pien CIP-GEGEVENS KONINKLIJKE BIBLIOTHEEK DEN HAAG Tanke, Richardus Hend. ^kus Johannes Experimental studies of neutron irradiated uranium dioxide at high temperatures / Richardus Hendrikus Johannes Tanke. - Arnhem : KEMA. - 111. Thesis Utrecht. - With index, ref. - With summary in Dutch. ISBN 90-353-1023-3 pbk. ISBN 90-353-1024-1 geb. SISO 538 UDC [539.125.06:546.791-31:536.45](043.3) Subject headings: uranium dioxide / gamma tomography. Contents Summary and outline 7 Outline of this thesis 9 Chapter 1: Introduction 11 The safety issue during the development of nuclear power 11 Experimental programmes on the release of fission products 15 from overheated fuel The KEMA source term experiment 17 References 18 Chapter 2: Experimental equipment and laboratory facilities 21 The hot cell 23 Measurement equipment for the evaporation experiment
    [Show full text]
  • Nuclear Fuel Cycle
    Quality and Standards for your converted material Hex Business at Springfields • Westinghouse will deliver on its promise of providing quality assurance for your company. • We pride ourselves on our engineering expertise and technical ability working to international standards including ISO 9001, 14001 and 18001. • Springfields has a long and successful history in the fuel cycle with over 40 years expertise in Hex conversion. Natural Mining Enrichment Fuel Fabrication Milling Nuclear Conversion Fuel Cycle Power Plant Reprocessing Electricity High Level Waste Storage Spent Fuel Storage www.westinghousenuclear.com Hex Business Rotary Kiln Plant Hex Plant Modern nuclear reactor designs such as Pressurised Water and Light Water Reactors To Atmosphere Uranium Trioxide UO – UF Conversion UF – UF Conversion UO 3 4 4 6 need nuclear fuel, made from Uranium Dioxide. 3 UO3 is transported to Kiln Plant in drums where it Uranium Hexafluoride is produced by the reaction UO To Atmosphere 3 is transferred onto a conveyor system. From here of UF with elemental gaseous fluorine in a Uranium Hexafluoride (UF6 ) is an essential intermediate product used in the Feed Hopper 4 Run Off Discharge the drums are delivered into the process via a fluidised bed reactor at 475°C. The fluorine manufacture of Uranium Dioxide fuels. Water UF6 mechanised tipping system. is produced by the electrolysis of anhydrous Condensers Hydration is the first stage in the treatment of hydrofluoric acid (AHF) in a potassium bi-fluoride Vac Back Air UO for the production of UF . electrolyte. 3 4 Secondary Meeting World Demands Advanced Technology Water Glycol Hydrator Heating & Cooling UO2 Transfer AHF Primary The Springfields Hex facilities are capable of Both plants demonstrate proven and state- To Scrubbers Hopper To Scrubbers H2 Transport Rotary Off-Gas Cylinder producing up to 5,500 tU of (UF6) annually, of-the-art technology.
    [Show full text]
  • The Carbon Reduction of Uranium Oxide" (1964)
    View metadata, citation and similar papers at core.ac.uk brought to you by CORE provided by Digital Repository @ Iowa State University Ames Laboratory Technical Reports Ames Laboratory 10-1964 The ac rbon reduction of uranium oxide H. A. Wilhelm Iowa State University Follow this and additional works at: http://lib.dr.iastate.edu/ameslab_isreports Part of the Ceramic Materials Commons, and the Metallurgy Commons Recommended Citation Wilhelm, H. A., "The carbon reduction of uranium oxide" (1964). Ames Laboratory Technical Reports. 86. http://lib.dr.iastate.edu/ameslab_isreports/86 This Report is brought to you for free and open access by the Ames Laboratory at Iowa State University Digital Repository. It has been accepted for inclusion in Ames Laboratory Technical Reports by an authorized administrator of Iowa State University Digital Repository. For more information, please contact [email protected]. The ac rbon reduction of uranium oxide Abstract The preparation of uranium metal directly from its oxides has been studied by a number of investigators. Various approaches have been pursued in these studies. Some attempts to produce the metal by electrolysis of the oxide in fused salt have been made; however, the reductions of the oxides by another metal and by carbon have attracted most attention. Although no entirely satisfactory method has yet been developed for preparation of uranium metal directly from oxide, the ready availability of high purity oxides makes such a process of interest. An investigation dealing with an approach to this problem is reported here. Disciplines Ceramic Materials | Metallurgy This report is available at Iowa State University Digital Repository: http://lib.dr.iastate.edu/ameslab_isreports/86 IS-1023 IOWA STATE UNIVERSITY THE CARBON REDUCTION OF URANIUM OXIDE by H.
    [Show full text]
  • Thermal Diffusivity of Uranium Dioxide Fuel Pellets with Addition of Beryllium Oxide by the Laser Flash Method
    XXI IMEKO World Congress “Measurement in Research and Industry” August 30 September 4, 2015, Prague, Czech Republic THERMAL DIFFUSIVITY OF URANIUM DIOXIDE FUEL PELLETS WITH ADDITION OF BERYLLIUM OXIDE BY THE LASER FLASH METHOD Ricardo A. N. Ferreira1, Denise M. Camarano2, Wilmar B. Ferraz3, Pablo A.Grossi4, Fábio A. Mansur5 1,2,3,4,5Centro de Desenvolvimento da Tecnologia Nuclear, Comissão Nacional de Energia Nuclear, Belo Horizonte, Brazil, [email protected] [email protected] [email protected] [email protected] [email protected] Abstract A research program is in progress at CDTN – Thus, an assess of improvements can be performed by Centro de Desenvolvimento da Tecnologia Nuclear, aiming comparison of the thermal diffusivity of pure uranium the development of fuel pellets made from a mixture of dioxide samples, and samples of uranium dioxide (UO2) uranium dioxide microspheres or powder and beryllium with addition of 2wt.% of beryllium oxide, produced with oxide powder for increasing the thermal conductivity of the the same batch of kernels and powder. This paper nuclear fuel. This type of fuel promises to be safer than complements the previous experiments and presents the current fuels, improving the performance of the reactor. In obtained results of thermal diffusivity at room temperature previous studies conducted at CDTN, fuel pellets were of these fuel pellets. manufactured and tested considering the following weight percentages: 1, 2, 3, 5, 7 and 14wt.% of beryllium oxide 2. METHODOLOGY (BeO) in mixture with microspheres of UO2. Continuing this project, new uranium dioxide fuel pellets with addition of 2.1. Fuel Pellets Fabrication with Uranium Dioxide 2wt.% of BeO were manufactured and tested.
    [Show full text]
  • Kinetic Study of the Oxidative Dissolution of UO2 in Aqueous Carbonate Media
    8188 Ind. Eng. Chem. Res. 2004, 43, 8188-8193 Kinetic Study of the Oxidative Dissolution of UO2 in Aqueous Carbonate Media Shane M. Peper,† Lia F. Brodnax,† Stephanie E. Field,† Ralph A. Zehnder,† Scott N. Valdez,‡ and Wolfgang H. Runde*,† Chemistry and Nuclear Materials Technology Divisions, Los Alamos National Laboratory, Los Alamos, New Mexico 87545 The oxidative dissolution of uranium(IV) dioxide powder at room temperature in aqueous carbonate media has been investigated. Kinetic studies evaluating the efficacy of various oxidants, including K2S2O8, NaOCl, and H2O2, for dissolving UO2 in alkaline solution have been performed, with H2O2 exhibiting the most rapid initial dissolution at 0.1 M oxidant concentrations. This result is due in part to the ability of peroxide to act as both an oxidant and a ligand under alkaline conditions. A spectrophotometric titration was used to confirm peroxide coordination to the U(VI) metal center. The disappearance of characteristic absorbance maxima associated 4- with UO2(CO3)3 (e.g., 448.5 nm) and a subsequent change in solution coloration upon titration with hydrogen peroxide indicated a change in speciation. Optimization of the hydrogen peroxide concentration indicated that the initial rate of uranium oxidation increased with increasing peroxide concentration, with a maximum reaction rate estimated at about 0.9 M peroxide. In addition, the effects of both the carbonate countercation and the carbonate concentration were also studied. It was determined that for 40 mg UO2 0.5MNa2CO3 was the most propitious choice, exhibiting both a high initial dissolution rate and the highest UO2 dissolution capacity among the systems studied.
    [Show full text]
  • Sintering and Thermal Behavior of Uranium Dioxide In
    SINTERING AND THERMAL BEHAVIOR OF URANIUM DIOXIDE IN BERYLLIUM OXIDE MATRIX A Thesis by CHAD BENJAMIN GARCIA Submitted to the Office of Graduate and Professional Studies of Texas A&M University in partial fulfillment of the requirements for the degree of MASTER OF SCIENCE Chair of Committee, Sean M. McDeavitt Committee Members, Jean C. Ragusa Miladin Radovic Head of Department, Yassin A. Hassan May 2014 Major Subject: Nuclear Engineering Copyright 2013 Chad Benjamin Garcia ABSTRACT The primary fuel used in light water nuclear reactors (LWRs) is uranium dioxide (UO2), which has a low thermal conductivity that causes a large thermal gradient across the fuel pin during operation. One proposed method to improve the thermal conductivity of the fuel is to insert a thermally conductive additive such as beryllium oxide (BeO) or other ceramic materials into the fuel structure. This study is focused on a particular fuel design developed at Purdue University wherein large UO2 microspheres are dispersed within a continuous BeO matrix. The BeO has a relatively high thermal conductivity for an oxide material and the interconnected matrix is intended to enable higher heat removal from the fuel. Therefore it is of interests to characterize the effective thermal conductivity of this UO2-BeO fuel concept and consider variations of this property with BeO composition of 2.5, 5, 7.5, and 10 volume % BeO. As part of this study, the basic pellet manufacturing procedures were improved over previous work to create samples suitable for characterization that did not crack or have a large volume of porosity. The pellet pressing and sintering methods were reproduced and then modified; densification measurements were performed to track the pellet status prior to, during, and after sintering using geometric measurements, immersion mass-based measurements, and LVDT dilatometery.
    [Show full text]
  • 'Ray Studies of Uraniumâ•filanthanumâ•'Mixed Oxides
    J. Am. Ceram. Soc., 98 [7] 2278–2285 (2015) DOI: 10.1111/jace.13559 © 2015 The Authors. Journal of the American Ceramic Society published by Wiley Periodicals, Inc. on behalf of American Ceramic Society (ACERS) This is an open access article under the terms of the Creative Commons Attribution-NonCommercial-NoDerivs License, which permits use and Journal distribution in any medium, provided the original work is properly cited, the use is non-commercial and no modifications or adaptations are made. Raman and X-ray Studies of Uranium–Lanthanum-Mixed Oxides Before and After Air Oxidation Zeynep Talip,† Thierry Wiss, Philippe E. Raison, Jerome^ Paillier, Dario Manara, Joseph Somers, and Rudy J.M. Konings European Commission, Joint Research Centre, Institute for Transuranium Elements, P. O. Box 2340, Karlsruhe 76125, Germany UO2 samples doped with 6, 11, 22 mol% lanthanum were first part describes the La-doping effect on the UO2 structure. examined before and after air oxidation. To verify the forma- The second part describes the low-temperature air oxidation tion of uranium–lanthanum-mixed oxide solid solutions, powder effect on pure and La-doped UO2 samples. X-ray diffraction (XRD) analyses of the crystalline phases in the materials were carried out. The presence of oxygen vacan- cies in the La-doped UO2 samples was identified by Raman II. Experimental Procedure spectrometry. It was evidenced by changes induced in the (1) Samples Raman spectra by air oxidation. This latter was carried out La-doped UO2 samples with three different compositions (6, either by increasing the Raman laser power or by thermally 11, 22 mol%) were prepared by a sol–gel technique.16 Trans- treating the samples at 500 K for 370 h.
    [Show full text]