The Manufacture of Uranium Dioxide Fuel in Pellet Form

Total Page:16

File Type:pdf, Size:1020Kb

The Manufacture of Uranium Dioxide Fuel in Pellet Form N\ THE MANUFACTURE OF URANIUM DIOXIDE FUEL IN PELLET FORM Arnold Blum Iscar Ltd , Nahariya Introduction Commercial nuclear fuel for conventional pressurized water reactors (PWR) is being produced in tonnage quantities on the basis of low enri- ched uranium oxide (UO ), i.e., UO in which the uranium is enriched to 5% or less in the fissionable isotope U-235. The bulk of the fuel proces- sed is of an enrichment between 2-3% U-235, While the fuel is generally drawn from the enrichment plants in the form of UFR, there are a variety of processes by which it is converted to uranium oxide. The method of conversion has a significant bearing on the properties of the uranium oxide powder which serves as the stating mate- rial for pellet fabrication This problem is being dealt with by another, and will not ba treated here, A fossil-fuel furnace is principally designed to generate heat, but in the process of doing so, also produces fay-products in the form of so- lid residues and gaseous effluents. A pressurized water reactor is ana- logous in its mode of operation. One of the solid residues of the fuel is the metal plutonium, which can be re-cycled and used in subsequent fuel cores. The fabrication of plutonium containing fuel, irrespective of its use, falls outside the scope of this paper. The fabrication of low enriched uranium oxide pellets is in essence a process for generating ceramic bodies, but the fabrication processes are complicated by problems of nuclear safety, security, health & safety, high intrinsic value, enrichment control, scrap recovery, accountability, effluent control and environmental protection, and the high sensitivity of the fuel to even minor contaminants, which are of no consequence in - 2 - conventional ceramics. The problems peculiar to the production of nuclear fuel result in a high overhead operation, with a fixed cost unheard of in most other industrial operations. The basic process, and some of the problems alluded to abova, will be dealt with in the following parts of this paper. The Basic Process 1. Powder Since the mechanism by which UO^ powder is converted to fuel pel- lets is based on the application of pressure and heat, the required magnitude of these, depends in large measure on the intrinsic surface energy of the powder. On the National Archives in Washington, D.C. is found the inscription: "The Past is Prologue". The truth of this state- ment can nowhere's be demonstrated as effectively as in the fabrication of UO, pellets. The impact of powder properties on the economics of fuel fabrica- tion is overwhelming, in regard to the reguired compacting pressures, sintering temperatures and sinterinq times, and the resulting density and integrity of fuel pellets which have a direct bearing on product yield. At the time of my last involvement in the fabrication of pellets, about one year ago, there was still raging a controversy among fuel designers as to the most desirable structure of sintered fuel.. Since in the fission process, fission gases are generated, these must be acco- modaterl inside the fuel pellets if these are not to disintegrate into dust under the influence of internal gas pressure. On one side of the argument were lined up the advocates of moderate density pellets (92 - 93% of theoretical) characterized by a microstructure with a high density matrix interspersed by well rounded, uniformly distributed pores. - 3 - Such a structure, because of the spherical nature of the porosity, would not be expected to shrink significantly under the influence of the hiqh temperatures prevailing inside the fuel under reactor operating condi- tions . On tho other side of the controversy were the advocates of high den- sity pellets (96% of theoretical and higher) who argued that on high burn- up, fuel shrinkage was bound to occur even if the pores were well rounded in medium density structures, and such shrinkage would tend toward fuel clad failure by tube buckling. Some such failures had just been discovered in operating reactors, and these reactors shut-down by order of the U.S. A.E.C. The outcome of this controversy will h»ve a decided impact on the types of powders required. High density pellets require powders with a specific surface of 3 rn /gram and up and Fisher Sub sieve Sizes in the range of 0.8 - 1.1 microns. Medium density pellets require lower specific surfaces, about 2 2.5 - 2.8 m /gram and Fisher Sub Sieve Sizes of 1.2 - 1.5 microns. There are, of course, other means of affecting the density of fuel pellets, such as the magnitude and kind of the lubricant addition and the mariner in which it is applied; the magnitude of U,0o additions, used as the preferred method of recycling sintered scrap; the compacting pressure; and the sintering temperature. But the predominant factor in achieving a desired sintered fuel pellet structure lies in the quality of the starting powder. Attempts were made in correlating the particle size distribution with the sintering behavior of powders, using 3 different types of instruments. None of the results were successful. - 4 - Another important quality attribute of UO powders is that of the 0/U ratio. Since UO is not the hiohest oxide of uranium, it can readily oxidize in the'finely'divided state. A certain amount of re-oxidation is almost inevitable as UO powder exits from the reduction step and is auto- natically conveyed through mechanical treatments to the weighing and blen- ding station. Dry ice is used to provide a CO blanket over exposed UO and also to extinguish fires which sometimes start spontaneously in expo- sed containers. U/0 ratios greater than 2.15 are not acceptable for pellet powder. Great care is exercised in maintaining the chemical and isotopic purity of UO- powders. Since mechanical conveyances are used in the preparation of the powder, there is the danger of contamination by abrasion of the contacting surfaces, principally by iron group metals. If is for this reason that 400 series stainless steel is preferred as material of cons- truction of contacting surfaces. This series of stainless, being magnetic, can be removed by magnetic separation. Neutron absorbers, such as cadmium, used as plating for steel faste- ners, and silver, used as a constituent of braze joints, are carefully excluded from powder and pellet facilities. So are boron, and the rare earths. The powders are analyzed for a lona list of elements, each speci- fied as to the permissible maximum concentration allowed. Isotopic analysis by raass-spectograph is required for each powder and pellet lot. In a plant processing a diversity of enrichments, a thorough "isotopic" cleanup between different enrichments is a major effort. Ins- pection and release by Quality Control is required before the facility may be used for a new enrichment. - 5 - 2. Press Feod Preparation I all automatic compactino presses the reproducibility of green density, weight and heiaht of a compact depends on reproducibly volume- tric die filling. In order to assure this, the powder must be rendered flowable in a reproducible manner. This requires the addition of organic lubricants which tend to hold tooether aggregates of primary particles to form spheres or granules which flow into the pressing die. If the spheres are imperfect, or fines are present in the press feed, they tend to cause improper die filling by "bridging", and the weight of material delivered to the dies fluctuates within excessive limits. Conventional press-feed preparation uses liaht compaction of the raw powder into low density compacts which are pushed through a screen and so broken up into granules. These granules are blended with a small percen- tage of an organic binder and rounded off into balss during the blending operation, so as to make them free-flowing. Some processors screen the blended material in order to remove fines and large balls, which impede flow. More advanced means of accomplishiRo this parpose are by the use of a "spray-drier". In such equipment a slurry composed of raw powder, an organic vehicle and the binder is pumped t.o a drying tower, where it is sprayed upward from a nozzle and the volatile vehicle is evaporated by a counter-current flow of hot, inert gas, causing the droplets of slurry to congeal into small balls or spheres which are then removed from the tower. Such press feed is more uniform, softer and freer of contamination, since it has not been exposed to abrasion as in the case of mechanical granulation. - 6 - However, nuclear safety aspects of the spray-drying techniaue have to be carefully engineered, due to the exposure of the UO-to fair-sized volumes of hydrogenous vehicles. Yet other press-feed results directly from the powder making process, in which the primary aggregates are generated in spherical form. Such aggregates are free flowing without an intermediate processing step. Lu- bricant must then be added to the dip walls during the pressino operation, since the close fitting punches cannot operate without lubrication. 3. Pellet Pressing Si.nce a 3000 kg. lot of powdr;r is converted into several hundred thousand pellets depending on pellet size and density, the press must be capable of high productivity and the tooling must be highly abrasion resis- tant to survive a production run. Cemented tungsten carbide tooling is the- refore standard for UO' pellet production. To minimize pellet-cladding interaction, as fuel clad tubes shrink onto the fuel stack, it has been found advisable to chamfer fuel pellets. Furthermore as fuel center line temperatures tend to be far higher than periphereal temperatures, there is a non-uniform axial expansion of fuel pellets in an operating reactor.
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]
  • 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.
    [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]