DOCSLIB.ORG

Modeling of the Thermal Behavior of Irradiated Sphere-Pac Mixed Carbide Fuel

Total Page：16

File Type：pdf, Size：1020Kb

Download full-text PDF Read full-text
Modeling of the Thermal Behavior of Irradiated Sphere-Pac Mixed Carbide Fuel AN ABSTRACT OF THE THESIS OF Maurice Ades for the degree of Doctor of Philosophy in Nuclear Engineering presented on February 28, 1979 Title: Modeling of the Thermal Behavior of Irradiated Sphere-Pac Mixed Carbide Fuel Redacted for Privacy Abstract Approved: ft f K. L. Peddicord Several fuels are being investigated for fast breeder reactor applications. One promising concept is sphere-pac mixed carbide fuel. As a part of the development of this fuel concept it is necessary to determine the behavior of the irradiatedfuel for safe and reliable operation on a sound economic basis. It is with the thermal aspect of this behavior that the current investigation is concerned. With a view to predicting the overall thermal behavior of the fuel pin during irradiation, theoretical models forvarious phenomena have been developed so that the various interrelated thermal components can be determined. These components include the thermal conductivity of the fuel in its initialconfiguration and during restructuring, the restructuring boundary, thetemperature distribution, the extent of restructuring due tosintering, grain growth and porosity redistribution, fuelswelling and gas release, and the behavior of the fill gas inthe free volume of the fuel pin. A two-dimensional unit cell model is developed toevaluate the effective conductivity of the fuel in itsinitial configuration. By implementing descriptions of sintering mechanisms on this model, the effective conductivity of the fuel during restructuring is evaluated. The restructuring boundary at each axial section is determined based on a thermal conductivity criteria. Under certain conditions the initial fuel configuration in the inner fuel regions is lost and is replaced by a porous pellet-shaped region. In the outer regions, however, fuel restructuring is not completed. The radial temperature distribution is evaluated by finite differences at a discrete number of axial sections. The local heat source has previously been computed foreach section. Allowance is made for axial asymetry as well as power histories. The porosity redistribution in the inner fuel regions which have completed restructuring is evaluated by using a model describing the conser- vation of pores migrating up the thermal gradient. Swelling and gas release are modeled allowing for bubble nucleation, bubble and pore growth and re-solution of gas atoms. Gas release occurs through diffusion of individual gas atomsin the fuel matrix and follows the subsequent buildup ofcontiguous inter- granular bubbles. Allowance is made for changes in the fuel micro- structure due to restructuring andthermally-activated grain growth. Following gas release, the temperature and pressureof the free gas are evaluated. The models describing the various thermalbehavior components were embodied into the computer code SPECKLE-I to predict the thermal response of the irradiated fuel pin. Results computed from SPECKLE -I indicate that the effective fuel conductivity increases with temperature, fill gas pressure and with the extent of sintering occuring between the fuel spheres. As gas release occurs, the effective conductivity is altered by the fission gas and the pressure buildup in the free volume. The fuel temperature, initially high, decreases significantly due to restructuring and is affected by the compensating effects of gas release and pressure buildup. The extent of restructuring and porosity redistribution increases with the fuel power. Fuel swelling and gas release are essentially controlled by the fuel microstructure and porosity. The results indicate that the extent of fuel restructuring affects swelling and gas release. As gas release occurs, the pressure of the fill gas increases significantly, its temperatureremaining almost constant. Results from SPECKLE-I were compared with several fuelirradia- tions. The agreement obtained between predicted and measuredthermal components suggests that SPECKLE-I is a valuable first stepfor evaluating the thermal response of irradiated sphere-pacmixed carbide fuel. The performance of this code, however, can beimproved by further model development and experimentalstudies of phenomena not fully characterized. MODELING OF THE THERMAL BEHAVIOR OF IRRADIATED SPHERE-PAC MIXED CARBIDE FUEL by Maurice Ades A THESIS submitted to Oregon State University in partial fulfillment of the requirements for the degree of Doctor of Philosophy Commencement June 1979 APPROVED: Redacted for Privacy Professor of Nuclear Engineering In Charge of Major Redacted for Privacy Head of Department of Nuclear Engineering Redacted for Privacy Dean of Graduate School Date thesis is presented February 28, 1979 Typed by Mary Bauman for Maurice Ades ACKNOWLEDGEMENT The present investigation was sponsored by EIR, the Swiss Federal Institute for Reactor Research, Wuerenlingen, Switzerland. Funding for my research by EIR is gratefully acknowledged. I would like to thank my major professor, Dr. K.L. Peddicord for his continued support and guidance during the course of this work. Special thanks are also due to R.W. Stratton, Director of the EIR Fuel Development Program, for his interest and criticism during the realization of the present investigation. I also would like to thank Drs. H. Blank, Hj. Matzke and C. Sari of the Euratom Transuranium Institute, Karlsruhe, Germany, for their appropriate comments on modeling of carbide fuel. Thanks are also due to Dr. J.R. Matthews and Mr. R.D. Bellamy of AERE, Harwell, U.K., for fruitful discussions on modeling. I am also indebted to Dr. A.H. Robinson for his various comments on appropriate numerical methods. At last, but not at least, I am grateful to my spouse, Rebecca, for her encouragement and sacrifice without which this work would not have been possible. TABLE OF CONTENTS Page I. Introduction I.1. General Background 1 1.2. Modeling of the Behavior of Irradiated Pellet Shaped Mixed Oxides 6 1.3. Modeling of the Behavior of Irradiated Pellet Shaped Mixed Carbides 21 1.4. Sphere-Pac Fuel 1.4.1. Description 29 1.4.2. Modeling of Irradiated Sphere-Pac Mixed Carbide Fuels 35 1.5. Research Objectives 37 II. Analytical Description of the Models II.1. Effective Conductivity of Sphere-Pac Fuel 43 II.1.1. Introduction 43 11.1.2. The Model 44 11.1.3. Calculational Procedure 49 11.1.4. Effective Conductivity of the Fine Fraction Unit Cell 49 11.1.5. Effective Conductivity of the Binary Unit Cell 51 11.1.6. Allowance for Fission Gas Release 52 11.2. Effective Conductivity of Sphere-Pac Fuel Undergoing Initial Stage Restructuring 55 11.2.1. Introduction 55 11.2.2. Sintering 56 11.2.2.1. Definition 56 11.2.2.2. Mechanisms 59 11.2.2.3. Effect of Applied Stress - Hot Pressing 65 11.2.2.4. Dominant Sintering Mechanism: Sintering Diagram 66 11.2.3. Evaluation of the Effective Conductivity of Sphere-Pac Fuel Undergoing Sintering . 68 11.2.3.1. The Method 68 11.2.3.2. Restructuring Boundary: The MCB Method 70 11.2.3.3. Effective Conductivity for Finite Time Step Computations 71 11.2.3.4. Allowance for Pressure Buildup and Fission Gas Release 74 11.3. Heat Source 75 11.4. Temperature Distribution 79 11.5. Intermediate Stage Fuel Restructuring 84 Page 11.5.1. Introduction 84 11.5.2. Porosity Redistribution 86 11.5.2.1. Introduction 86 11.5.2.2. The Model 91 11.5.3. Grain Growth 99 11.5.3.1. Introduction 99 11.5.3.2. The Model 101 11.6. Fuel Swelling and Gas Release 105 11.6.1. Introduction 105 11.6.2. Nucleation of Fission Gas Bubbles 106 11.6.3. Re-Solution 111 11.6.4. Allowance for Fuel Porosity 114 11.6.5. The Model 118 11.6.6. Method of Solution 122 11.6.7. Grain Boundary Bubbles 126 11.6.8. Volume Swelling 131 11.6.8.1. Volume Swelling Due to Fission Gas 132 11.6.8.2. Volume Swelling Due to Solid Fission Products 134 11.6.8.3. Total Volume Swelling 137 11.6.9. Gas Release 142 11.6.9.1. Gas Release From a Fuel Grain . 142 11.6.9.2. Total Gas Release 144 11.7. Free Gas Conditions 148 III. Computer Simulation of the Thermal Behavior of Irradiated Sphere-Pac Fuel III.1. Introduction 151 111.2. SPECKLE-I 151 IV. Results and Discussions IV.1. Application of the Models 158 IV.2. Effective Conductivity of Sphere-Pac Fuel 159 IV.3. Effective Conductivity of Sphere-Pac Fuel Undergoing Sintering 165 IV.3.1. Sintering Mechanisms 165 IV.3.2. Effective Conductivity of Sphere-Pac Fuel During Initial Stage Restructuring 175 IV.3.3. Allowance for Gas Pressure and Fission Gas Release 182 187 IV.4. Temperature Distribution and RestructuringBoundary IV.4.1. Computational Results 187 IV.4.2. Effect of Applied Stress 193 IV.4.3. Comparison with Experiments 197 Page IV.5. Porosity Redistribution 198 IV.5.1. Computational Results 198 IV.5.2. Comparison with Experiments 213 IV.6. Fuel Swelling and Gas Release 218 IV.6.1. Preliminary Results 218 IV.6.2. Computational Results 222 IV.7. Free Gas Conditions and Overall Thermal Behavior . 227 IV.8. Limitations of the Models 243 IV.8.1. Effective Conductivity of Sphere-Pac Fuel 244 IV.8.2. Effective Conductivity of Sphere-Pac Fuel Undergoing Sintering 245 IV.8.3. Heat Source 246 IV.8.4. Temperature Distribution 246 IV.8.5. Porosity Redistribution 247 IV.8.6. Grain Growth 249 IV.8.7. Fuel Swelling and Gas Release 250 V. Conclusions and Recommendations V.I. Conclusions 254 V.2. Recommendations for Future Theoretical Studies . 264 V.3. Recommendations for Future Experimental Studies . 267 References 271 Appendix 282 LIST OF ILLUSTRATIONS Figure Page 1. Schematic Representation of Pellet-Shaped and Sphere-Pac Fuel Before and After Irradiation 31 2. Fuel Microstructure and Porosity Distribution in Pelleted (U08Pu02)C Fuel 33 3. Conceptual Flow Chart for a Computer Program to Simulate the Behavior of an Irradiated Sphere-Pac Carbide Fuel Pin 38 4.
Load more

Recommended publications
  • Uranium Oxide Aerosol Transport in Porous Graphite
    PNNL-21014 Prepared for the U.S. Department of Energy under Contract DE-AC05-76RL01830 Uranium Oxide Aerosol Transport in Porous Graphite J Blanchard C Brown DC Gerlach C Lovin RD Scheele CH Delegard ML Stewart A Zelenyuk BD Reid EC Buck PA Gauglitz BJ Riley LM Bagaasen CA Burns January 2012 PNNL-XXXXX PNNL-21014 Uranium Oxide Aerosol Transport in Porous Graphite J Blanchard C Brown DC Gerlach C Lovin RD Scheele CH Delegard ML Stewart A Zelenyuk BD Reid EC Buck PA Gauglitz BJ Riley LM Bagaasen CA Burns January 2012 Prepared for the U.S. Department of Energy under Contract DE-AC05-76RL01830 Pacific Northwest National Laboratory Richland, Washington 99352 PNNL-21014 Summary Conditions exist in carbon dioxide (CO2) gas-cooled, graphite moderated reactors that might allow for gaseous transport of uranium oxide (UO2) particles into the graphite moderator. The transport of UO2 in the reactor coolant system, and subsequent deposition of this material in the graphite, is of interest due to the potential to influence the application of the Graphite Isotope Ratio Method (GIRM). GIRM was developed to validate the declared operation of graphite moderated reactors. Uranium impurities in nuclear grade graphite are one of several possible indicator elements that can be used for a GIRM assessment. During fuel failures, uranium metal in the fuel is exposed to the CO2 coolant and oxidizes as either UO2 or U3O8. Measurements in adjacent fuel channels indicate that CO2 coolant readily flows through the porous graphite moderator blocks. Therefore, the potential exists for the coolant gas to transport UO2 particles, as an aerosol, into the porous graphite, thereby invalidating uranium as an indicator element.
    [Show full text]
  • 3 - 20 Progress in Fabrication of UC Ceramic Nuclear Fuels
    · 56 · IMP & HIRFL Annual Report 2018 formation of such carbonyl complexes. Reference [1] Y. Wang, S. W. Cao, J. C. Zhang, et al., Physical Chemistry Chemical Physics, 21(2019)7147. [2] Y. Wang, Y. Wittwer, J. Zhang, et al., PSI Annual Report, (2019). 3 - 20 Progress in Fabrication of UC Ceramic Nuclear Fuels Tian Wei and Qin Zhi Uranium dioxide nuclear fuel has been widely used in pressurized water reactors (PWR), because of its high melting point, isotropic expansion, excellent radiation behaviors and mechanical properties[1]. However, it is easy to embrittlement and the thermal conductivity is quite low. Compared with traditional uranium oxides nuclear fuel, the uranium carbide has extremely high hardness and no phase transition occurs in wide temperature range[2]. The thermal conductivity, density and uranium fraction of UC is 21.7 W/(m·K) (1 237 K), 13.63 g/cm3 and 95.2% respectively, which are much higher than that of UO2. Uranium carbide ceramic fuel has been considered as a potential candidate nuclear fuel for the next-generation fast-neutron reactors, especially for the Accelerator-Driven Systems (ADS). Based on the accelerator driven recycling used nuclear fuel, uranium carbide can recrystallize with plutonium (Pu) and minor actinides (MAs). Therefore uranium carbide is chosen as the fuel format in ADS. Uniform sized ceramic UC microspheres with a diameter of (675 10) µm were successfully fabricated by an improved microwave-assisted rapid internal gelation process combined with carbothermic reduction (Fig. 1). First of all, the nanoparticle carbon was dispersed into the HMUR stock solution, and the C-UO3·2 H2O gelled microspheres were prepared using an improved microwave-assisted internal gelation process without cooling the initial stock solutions.
    [Show full text]
  • Appendix a of Final Environmental Impact Statement for a Geologic Repository for the Disposal of Spent Nuclear Fuel and High-Lev
    Appendix A Inventory and Characteristics of Spent Nuclear Fuel, High-Level Radioactive Waste, and Other Materials Inventory and Characteristics of Spent Nuclear Fuel, High-Level Radioactive Waste, and Other Materials TABLE OF CONTENTS Section Page A. Inventory and Characteristics of Spent Nuclear Fuel, High-Level Radioactive Waste, and Other Materials ................................................................................................................................. A-1 A.1 Introduction .............................................................................................................................. A-1 A.1.1 Inventory Data Summary .................................................................................................... A-2 A.1.1.1 Sources ......................................................................................................................... A-2 A.1.1.2 Present Storage and Generation Status ........................................................................ A-4 A.1.1.3 Final Waste Form ......................................................................................................... A-6 A.1.1.4 Waste Characteristics ................................................................................................... A-6 A.1.1.4.1 Mass and Volume ................................................................................................. A-6 A.1.1.4.2 Radionuclide Inventories ...................................................................................... A-8 A.1.1.4.3
    [Show full text]
  • Preparation of Uranium Carbonitride by Reaction of UC with Ammonia
    Journal of NUCLEARSCIENCE and TECHNOLOGY,5[8], p. 414~418 (August 1968). Preparation of Uranium Carbonitride by Reaction of UC with Ammonia Yumi AKIMOTO* and Koji TANAKA* Received January 24, 1968 Revised April 11, 1968 With the view to establishing a method of directly converting uranium carbide into uranium carbonitride and hydrocarbons, an attempt has been made to induce reaction between UC and am- monia under various temperatures from 25d to 600dc and pressures from 1 to 1,500 kg/cm2. The reaction aimed at was realized to the extent of practical significance under pressures exceeding 500 kg/cm2 at 450dc and exceeding 250 kg/cm2 at 500dc. The hydrocarbons produced thereby were found to be mainly methane, indicating that the formula of the predominant reaction was UC+NH3 -> UN2-x+CH4+H2. Upon heating the powdery fine black product to 1,800dc in vacuo, unexpectedly marked sinter- ing was found to occur, resulting in dense uranium monocarbonitride without any compaction pre- treatment. carbon is also reported. I. INTRODUCTION Another commonly known nitriding reac- Uranium mononitride (UN), and uranium tion is that between uranium carbides and carbonitride (UC1-xNx) have attracted much nitrogen, to result in a pulverized mixture of attention as possible fuel materials with quali- higher uranium nitrides and carbon(6)(7): ties surpassing uranium carbide (UC) with 2UC+3/2N2=U2N3+2C. (2) respect to chemical stability and simplicity of stoichiometric control. From the standpoint This reaction also fails to serve the purpose of fuel cost however, UN and UC1-xNx still of producing UN or UC1-xNx because carbon, yield to UC because the process at present once produced, cannot easily be separated available for the preparation of pure UN and from the nitride.
    [Show full text]
  • Mathematical Model of the Oxidation of a Uranium Carbide Fuel Pellet Including an Adherent Product Layer
    This is a repository copy of Mathematical Model of the Oxidation of a Uranium Carbide Fuel Pellet Including an Adherent Product Layer. White Rose Research Online URL for this paper: http://eprints.whiterose.ac.uk/110903/ Version: Accepted Version Article: Shepherd, JS, Fairweather, M, Hanson, BC orcid.org/0000-0002-1720-1656 et al. (1 more author) (2017) Mathematical Model of the Oxidation of a Uranium Carbide Fuel Pellet Including an Adherent Product Layer. Applied Mathematical Modelling, 45. pp. 784-801. ISSN 0307-904X https://doi.org/10.1016/j.apm.2017.01.041 © 2017 Elsevier Inc. This manuscript version is made available under the CC-BY-NC-ND 4.0 license http://creativecommons.org/licenses/by-nc-nd/4.0/ Reuse Unless indicated otherwise, fulltext items are protected by copyright with all rights reserved. The copyright exception in section 29 of the Copyright, Designs and Patents Act 1988 allows the making of a single copy solely for the purpose of non-commercial research or private study within the limits of fair dealing. The publisher or other rights-holder may allow further reproduction and re-use of this version - refer to the White Rose Research Online record for this item. Where records identify the publisher as the copyright holder, users can verify any specific terms of use on the publisher’s website. Takedown If you consider content in White Rose Research Online to be in breach of UK law, please notify us by emailing [email protected] including the URL of the record and the reason for the withdrawal request.
    [Show full text]
  • Tlju^Tt^ PROPERTIES of FUELS for HIGH-TEMPERATURE REACTOR
    Tlju^tt^ PROPERTIES OF FUELS FOR HIGH-TEMPERATURE REACTOR CONCEPTS BATTELLE MEMORIAL INSTITUTE 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, makes 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. CORRIGENDA BMI- 1598, PROPERTIES OF FUELS FOR HIGH-TEMPERATURE REACTOR CONCEPTS July 1, 1963 Please attach to report or make the following changes on the appropriate pages. Page Fuel Comparison Table: Substitute the following values for the thermal con­ ductivities for UO2 in cal/(sec)(cm)(C): 0. 026 at 20 C, 0. 008 at 600 C, 0. 005 at 1200 C. 11 Section C6: Rupture time :fo r 1600 F siloul d be 60. 2 ins tead of 60 2. 17 Section C6: Revise as follows; Stress for Creep Rate Shown, psi Rupture Data Temperature, 0.001 Per 0.01 Per 0.
    [Show full text]
  • Synthesis and Optimization of the Sintering Kinetics of Actinide Nitrides
    ANNUAL PROGRESS REPORT Project Title: Synthesis and Optimization of the Sintering Kinetics of Actinide Nitrides Covering Period: July 2005 – December 2008 Date of Report: 03/31/2009 Recipient: Boise State University Award Number: DE-FC07 – 05ID14650 Subcontractors: University of Florida Other Partners: None Contact(s): Darryl P. Butt, 208-426-1054, [email protected] Boise State University 1910 University Drive Boise, ID 83725-2075 Project Team: John Carmack, INL (not funded); Ken McClellan, LANL (not funded); Ron Baney, UF (funded); James Tulenko, UF (funded); Tim Gerhardt, UF (funded); Le Song, UF (funded); Ravi Kumar Vasudevin, UF (funded); Brian M. Marx, BSU (funded); A. S. Hamdy (not funded); Patrick Callahan, BSU (funded); Brian Jaques, BSU (funded); Dan Osterberg BSU (funded). Authors: Darryl P. Butt and Brian Jaques i TABLE OF CONTENTS TABLE OF CONTENTS................................................................................................................. i LIST OF TABLES......................................................................................................................... iii LIST OF FIGURES ....................................................................................................................... iv NOMENCLATURE AND ACRONYMS ................................................................................... xiv CHAPTER I: OBJECTIVE AND EXECUTIVE SUMMARY.....................................................16 1. Objective..............................................................................................................................16
    [Show full text]
  • LOS ALAMOS SCIENTIFIC LABORATORY of the UNIVERSITY of CALIFORNIA LOS ALAMOS NEW MEXICO REPORT WRITTEN: March 1959 REPORT DISTRIBUTED: August 27, 1959
    LEGAL NOTICE ThlB report was prepared as an account of Government sponsored work. Neither the United States, nor the Commission, nor any person acting on behalf of the Commission A. Makes any warranty or representation, expressed or Implied, with respect to the accu­ racy, completeness, or usefulness of the information contained In this report, or that the use of any InformaUon, apparatus, method, or process disclosed in this report may not Infringe privately own^ rights, or B. Assumes any liabilities with respect to the use of, or for damages resulting from the use of any information, a^iaratus, method, or ptocess disclosed la this report As used in the above, "person acting on behalf of the Commission" Includes any em­ ployee or contractor of the Commission, or employee of such contractor, to the extent that such employee or contractor of the Commission, or employee of such contractor prepares, disseminates, or provides access to, any Information pursuant to his employment or contract with the Commission, or his employment with such contractor. LA-2306 CHE MISTRY- - GENERAL (TID-4500, 14th edition) LOS ALAMOS SCIENTIFIC LABORATORY OF THE UNIVERSITY OF CALIFORNIA LOS ALAMOS NEW MEXICO REPORT WRITTEN: March 1959 REPORT DISTRIBUTED: August 27, 1959 THE ANALYSIS OF REFRACTORY BORIDES, CARBIDES, NITRIDES, AND SILICIDES by Owen H. Kriege This report expresses the opinions of the author or authors and does not necessarily reflect the opinions or views of the Los Alamos Scientific Laboratory. Contract W-7405-ENG, 36 with the U. S. Atomic Energy Commission - 1 - DISCLAIMER This report was prepared as an account of work sponsored by an agency of the United States Government.
    [Show full text]
  • New Materials in Nuclear Technology
    including about 140 from outside the United Kingdom. with many of these associations, as well as with In more recent years similar organizations have been numerous individual medical physicists. An Inter­ formed in many other parts of the world, notably in national Organization for Medical Physics, linking the Australia, Canada, the Netherlands, Scandinavia, and various national associations, was inaugurated at the the USA. The Agency has close and cordial relations beginning of this year. NEW MATERIALS IN NUCLEAR TECHNOLOGY A major objective of current re search in nuclear technology is to develop reactor materials that can withstand the effects of high temperature and intense radiations. The efficiency of a nuclear power station - more specifically, the efficiency of utilizing the heat for the production of electricity - is partly dependent on the temperature at which the reactor can be op­ erated, but the working temperature cannot be in­ creased to the point at which the fuel elements will fail. This has led to a search for non-metallic forms of nuclear fuel, because in general these have a higher melting point than metallic fuels. Oxides of uranium, of course, have been used as reactor fuel from the earliest days of nuclear technology, but considerable work is still being done to improve the production and fabrication of oxide fuels so that they may be able to stand up to the progressively exacting conditions to which they may be subjected in a reactor. Simul­ The IAEA Director General, Dr. Sigvard Eklund taneously, work is being carried on to develop other (middle), addressing the opening session of the forms of non-metallic fuel, such as carbides, nitrides Prague conference.
    [Show full text]
  • An Evaluation of the Mechanical Properties and Microstructure In
    An Evaluation of the Mechanical Properties and Microstructure in Uranium Dioxide Doped with Oxide Additives by Robert McDonald A Thesis Presented in Partial Fulfillment of the Requirements for the Degree Master of Science Approved November 2014 by the Graduate Supervisory Committee: Pedro Peralta, Chair Jagannathan Rajagopalan Kiran Solanki ARIZONA STATE UNIVERSITY December 2014 ABSTRACT The United States Department of Energy (DOE) has always held the safety and reliability of the nation’s nuclear reactor fleet as a top priority. Continual improvements and advancements in nuclear fuels have been instrumental in maximizing energy generation from nuclear power plants and minimizing waste. One aspect of the DOE Fuel Cycle Research and Development Advanced Fuels Campaign is to improve the mechanical properties of uranium dioxide (UO2) for nuclear fuel applications. In an effort to improve the performance of UO2, by increasing the fracture toughness and ductility, small quantities of oxide materials have been added to samples to act as dopants. The different dopants used in this study are: titanium dioxide, yttrium oxide, aluminum oxide, silicon dioxide, and chromium oxide. The effects of the individual dopants and some dopant combinations on the microstructure and mechanical properties are determined using indentation fracture experiments in tandem with scanning electron microscopy. Indentation fracture experiments are carried out at room temperature and at temperatures between 450 °C and 1160 °C. The results of this work find that doping with aluminosilicate produces the largest favorable change in the mechanical properties of UO2. This sample exhibits an increase in fracture toughness at room temperature without showing a change in yield strength at elevated temperatures.
    [Show full text]
  • Compatibility Tests of Molten Uranium with Tungsten and Tungsten-1.5
    NASA TECHNICAL NOTE _-5076 *o h 90 n z c 4 r/, 4 z COMPATIBILITY TESTS OF MOLTEN URANIUM WITH TUNGSTEN AND TUNGSTEN- 1.5 PERCENT HAFNIUM by Philip D. Tukkzcnen Lewis Reseurch Center i 1; CZeveZund, Ohio *. NATIONAL AERONAUTICS AND SPACE ADMINISTRATION WASHINGTON, D. C. MARCH 1969 TECH LIBRARY KAFB, NM OL31917 NASA TN D-5076 COMPATIBILITY TESTS OF MOLTEN URANIUM WITH TUNGSTEN AND TUNGSTEN - 1.5 PERCENT HAFNIUM By Philip D. Takkunen Lewis Research Center Cleveland, Ohio NATIONAL AERONAUT ICs AND SPACE ADMINISTRATION For sols by the Cleoringhouse for Federal Scientific ond Technical Information Springfield, Virginia 22151 - CFSTI price $3.00 . .. ABSTRACT Isothermal compatibility tests were conducted on sealed tungsten (W) and tungsten - 1. 5-weight-percent-hafnium (W-l.5Hf) capsules containing uranium (U) for 10 to 200 hours at 1200' and 1500° C. Extensive intergranular attack was observed in wrought W rod and in vapor-deposited W. Hafnium appeared to offer some promise in reducing intergranular attack on W. Single-crystal W appeared to be compatible with molten U although some evidence of mass transfer attack was found. In addition, U climbed up the sides of the single-crystal capsules and attacked grain boundaries in the closure weld. Cracks were formed in all the capsule walls during cooling after U had solidified. ii CONTENTS Page SUMMARY ....................................... 1 INTRODUCTION . 2 EXPERIMENTAL PROCEDURE . 3 Test Capsule Components . 3 Fabrication of Capsule Components . 5 Capsule Furnace Treatments and Evaluation . 9 RESULTS OF CAPSULE TESTS . 9 Tungsten Capsule Tests . 10 1200' C for 10 hours . 10 1200° C for 100 hours .
    [Show full text]
  • N Tj )~ J JJC1~ ....Z ~
    - 1 S'5YJOl{ 3~ . ~ ~f NASA TN D-1442 frr N -.:;fi .....-.:;fi qI N tJ )~ J JJC1~ ....z ~/ <(./) Z< TECHNICAL NOTE D-1442 COMPATIBILITY OF MOLTEN URANIUM DIOXIDE WITH FIVE REFRACTORY MATERlALS By William A. Sanders and Isadore L. Drell Lewis Research Center Cleveland, Ohio NATIONAL AERONAUTICS AND SPACE ADMINISTRATION WASHINGTON January 1963 NATIONAL AERONAUTICS AND SPACE ADMINISTRATION TECHNICAL NOTE D-1442 COMPATIBILITY OF MOLTEN URANIUM DIOXIDE WITH FIVE REFRACTORY MATERIALS By William A. Sanders and Isadore L. Drell SUMMARY An investigation of the compatibility of molten 'uraniUm dioxide /­ with tungsten, tantalum, tantalum diboride, tantaiilln carbide, and hafnium carbide was conducted to indicate combinations that'may be ' useful in nuclear reactors that are to operate' at temperatures 'above the melting point of uranium dioxide, 50000 F. The evaluations' were conducted at a test temperature of 52200 F for a period of 15 minutes in a high-purity argon atmosphere. One compatibility test of molten uranium dioxide with tungsten was conducted at 55000 F. The tests were' performed with the refractory material in the form of a crucible that contained a uranium,dioxide charge. Crucible cross sections were studied after testing, and the degree of reaction between uranium diox­ ide and the various crucible materials was judged by the use of X-ray diffraction analysis, metallographic examination, microhardness meas­ urements, and electron-beam microanalysis. All five materials reacted with molten uranium dioxide at ,the test temperature of 52200 F. For tungsten, the extent of reaction was . slight even at 55000 F. For tantalum, the reaction'was greater than but similar in nature to that involving tungsten.
    [Show full text]