High-Density Concrete with Ceramic Aggregate Based on Depleted Uranium Dioxide
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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. -
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. -
Depleted Uranium: a DOE Management Challenge
Depleted Uranium: A DOE Management Challenge October 1995 U.S. Department of Energy Office of Environmental Management Office of Technology Development This report has been reproduced direcdy from die best available copy. Available to DOE and DOE Contractors from the Office of Scientific and Technical Information, P.O. Box 62, Oak Ridge, TN 37831; prices available from (615) 576-8401. Available to die pubfic from die U.S. Department of Commerce, Technology Administration, National Technical Information Service, Springfield, VA 22161, (703) 487-4650. Printed with soy ink on recycled paper DISCLAIMER This report was prepared as an account of work sponsored by an agency of the United States Government. Neither the United States Government nor any agency thereof, nor any of their employees, make any warranty, express or implied, or assumes any legal liability or responsibility for the accuracy, completeness, or usefulness of any information, apparatus, product, or process disclosed, or represents that its use would not infringe privately owned rights. Reference herein to any specific commercial product, process, or service by trade name, trademark, manufacturer, or otherwise does not necessarily constitute or imply its endorsement, recommendation, or favoring by the United States Government or any agency thereof. The views and opinions of authors expressed herein do not necessarily state or reflect those of the United States Government or any agency thereof. DISCLAIMER Portions of this document may be illegible in electronic image products. Images are produced from the best available original document. DOE/EM-0262 Depleted Uranium: A DOE Management Challenge October 1995 JSW O.STR.BUT.ON OF TH,S DOCUMENT » UNUM'TSD T U.S. -
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. -
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. -
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 -
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. -
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. -
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. -
Beneficial Uses of Depleted Uranium
BENEFICIAL USES OF DEPLETED URANIUM Colette Brown U.S. Department of Energy Germantown, Maryland 20871 Allen G. Croff and M. Jonathan Haire Chemical Technology Division Oak Ridge National Laboratory* Oak Ridge, Tennessee 37831-6180 Prepared for the Beneficial Re-Use ‘97 Conference Knoxville, Tennessee August 5–7, 1997 _________________________ *Managed by Lockheed Martin Energy Research Corp., under contract DE-AC05-96OR22464 for the U.S. Department of Energy. BENEFICIAL USES OF DEPLETED URANIUM1 Colette Brown Allen G. Croff and M. Jonathan Haire U.S. Department of Energy Oak Ridge National Laboratory2 19901 Germantown Rd. P.O. Box 2008 Germantown, Maryland 20871 Oak Ridge, Tennessee 37831-6180 (301) 903-5512 (423) 574-7141 INTRODUCTION Naturally occurring uranium contains 0.71 wt % 235U. In order for the uranium to be useful in most fission reactors, it must be enriched—the concentration of the fissile isotope 235U must be increased. Depleted uranium (DU) is a co-product of the processing of natural uranium to produce enriched uranium, and DU has a 235U concentration of <0.71 wt %. In the United States, essentially all of the DU inventory is in the chemical form of uranium hexafluoride (UF6) and is stored in large cylinders above ground. If this co-product material were to be declared surplus, converted to a stable oxide form, and disposed, the costs are estimated to be several billion dollars. Only small amounts of DU have at this time been beneficially reused. The U.S. Department of Energy (DOE) has begun the Beneficial Uses of DU Project to identify large-scale uses of DU and encourage its reuse for the primary purpose of potentially reducing the cost and expediting the disposition of the DU inventory. -
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. -
1 Low Ph Concrete for Use in the US High-Level Waste Repository: Part I Overview
1 Low pH Concrete for Use in the US High-Level Waste Repository: Part I Overview Leslie R. Dole and Catherine H. Mattus Nuclear Science Technology Division of Oak Ridge National Laboratory 1.1 Introduction The current baseline plan for the Yucca Mountain (YM) repository project (YMP) uses steel structures to stabilize the disposal drifts that are collectively over 65 kilometers in length (Figure 1-1). However using cementitious materials and application techniques, the potential exists to reduce the underground construction cost by 100’s of million of dollars, accelerate construction, and improve the repository’s performance. This development program was designed to meet the economic and engineering properties of the appropriate cementitious materials to build out these tunnels1. This work supposes that cementitious liners and inverts (the flat bedding placed in the drifts upon which the rails are laid) will be used ultimately in the second-generation repository design. This paper describes the required properties of Figure 1-1. Current YMP baseline underground tunnel YM-compatible cements build-out construction technology. and the program to develop and test materials for a suite of underground construction technologies. Specifically, this paper addresses mitigation of potential impacts of free calcium hydroxide in the concrete and grout binding matrices. Part II describes the formulation and testing in detail. 1.2 Background Based on several concerns described throughout the YMP document on the Evaluation of Alternative Materials for Emplacement Drift Ground Control (BCAA00000-01717-0200- 00013 REV 00)2, the wide use of cements is currently not included in the YM repository construction plans.