Citation
Feasibility of processing mixed carbide nuclear fuels for use in transmutation systems

Material Information

Title:
Feasibility of processing mixed carbide nuclear fuels for use in transmutation systems
Creator:
Carter, Thomas Clifford ( Dissertant )
Anghaie, Samin. ( Thesis advisor )
Dugan, Edward ( Reviewer )
Place of Publication:
Gainesville, Fla.
Publisher:
University of Florida
Publication Date:
Copyright Date:
2005
Language:
English

Subjects

Subjects / Keywords:
Carbides ( jstor )
Carbon ( jstor )
Density ( jstor )
Graphite ( jstor )
Nitrides ( jstor )
Sintering ( jstor )
Solid solutions ( jstor )
Transmutation ( jstor )
Uranium ( jstor )
X ray diffraction ( jstor )
Dissertations, Academic -- UF -- Nuclear and Radiological Engineering ( local )
Nuclear and Radiological Engineering thesis, M.S ( local )

Notes

Abstract:
The Advanced Fuel Cycle Initiative (AFCI) is seeking a fuel form to be used as a target material for the transmutation of actinides in either accelerator driven or reactor based fast spectrum systems. Currently funded research by the AFCI fuel group has been in the areas of oxide, nitride, metallic, and dispersion fuels. This work focuses on the processing of mixed actinide/refractory metal carbides as a potential fuel to be used in these systems. The objective of this work was to produce a low porosity, solid solution, single phase, mixed carbide while staying within parameters set by previous experience with processing minor actinides. Uranium was used as a surrogate for the actinides due to the constraint set by the available facilities. Ten samples of (U₀.₆₄,Zr₀.₃₆)C₀.₉₂ were processed by mixing powders, cold pressing at 180 and 300 MPa, and sintering for 10 hours at less than 2100 K in an induction furnace. The upper limit of sintering was set by the hypothetical americium loss that would occur if it had actually been used in the study. Consolidation was difficult to achieve without pressing the samples at approximately 15 MPa while sintering. The samples that were pressed during sintering all showed a solid solution and single phase structure. A density of greater than 90% of the theoretical density was achieved in one sample. Previous experience has not shown that a solid solution could be formed at temperatures that do not induce liquid phase sintering. This work has shown that the application of a moderate pressure during sintering helps for this solid solution formation to occur. ( , )
Subject:
actinides, americium, carbides, neptunium, nuclear, plutonium, transmutation
General Note:
Title from title page of source document.
General Note:
Document formatted into pages; contains 57 pages.
General Note:
Includes vita.
Thesis:
Thesis (M.S.)--University of Florida, 2005.
Bibliography:
Includes bibliographical references.
General Note:
Text (Electronic thesis) in PDF format.

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Source Institution:
University of Florida
Holding Location:
University of Florida
Rights Management:
Copyright Carter, Thomas Clifford. Permission granted to the University of Florida to digitize, archive and distribute this item for non-profit research and educational purposes. Any reuse of this item in excess of fair use or other copyright exemptions requires permission of the copyright holder.
Embargo Date:
7/30/2007
Resource Identifier:
71638217 ( OCLC )

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FEASIBILITY OF PROCESSING MIXED CARBIDE NUCLEAR FUELS FOR USE IN TRANSMUTATION SYSTEMS By THOMAS CLIFFORD CARTER A THESIS PRESENTED TO THE GRADUATE SCHOOL OF THE UNIVERSITY OF FLOR IDA IN PARTIAL FULFILLMENT OF THE REQUIREMENTS FOR THE DEGREE OF MASTER OF SCIENCE UNIVERSITY OF FLORIDA 2005

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Copyright 2005 by Thomas C. Carter

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This work is dedicated to my sister, Ren ee Carter, for her consta nt love and support.

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iv ACKNOWLEDGMENTS I would like to thank my family for thei r consistent support through the many years of my education. Also, I would like to tha nk my advisor, Dr. Samim Anghaie, without whom this work would not be possible, as well as the other member of my committee, Dr. Edward Dugan. Dr. Travis Knight also deserves special thanks for his advice and mentorship through the early stages of this wo rk. All experiments were performed at the High Temperature Materials Laboratory at the Innovative Nuclear Space Power and Propulsion Institute. Analysis was performed at the Major Analytical Instrumentation Center at the University of Fl orida. I would also like to thank the Department of Energy, Office of Nuclear Energy, Science and T echnology and the Advanced Nuclear Fuel Cycle Initiative for the funding of my studies.

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v TABLE OF CONTENTS page ACKNOWLEDGMENTS.................................................................................................iv LIST OF TABLES............................................................................................................vii LIST OF FIGURES.........................................................................................................viii ABSTRACT....................................................................................................................... ..x CHAPTER 1 INTRODUCTION........................................................................................................1 1.1 Background Information.........................................................................................1 1.1.1 Advanced Fuel Cycle Initiative Development.............................................2 1.1.2 Advanced Fuel Cycle Initiative Structure....................................................2 1.1.3 History of Mixed Carbide Nuclear Fuel.......................................................5 1.2 Purpose of Work.....................................................................................................6 1.2.1 Motivation....................................................................................................6 1.2.2 Objective.......................................................................................................6 2 METHODS AND MATERIALS..................................................................................8 2.1 Processing...............................................................................................................8 2.1.1 Powder Processing........................................................................................8 2.1.2 Cold Uniaxial Pressing.................................................................................9 2.1.3 Sintering and Hot Press..............................................................................11 2.2 Characterization....................................................................................................14 2.2.1 Density Measurements...............................................................................14 2.2.2 Scanning Electron Microscopy...................................................................15 2.2.3 X-Ray Diffraction.......................................................................................15 2.2.4 Carbon Determination................................................................................15 3 RESULTS.................................................................................................................. .16 3.1 Consolidation and Density....................................................................................16 3.2 Scanning Electron Microscopy Analysis..............................................................17 3.3 X-Ray Diffraction.................................................................................................19 3.4 Carbon Determination..........................................................................................19

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vi 4 DISCUSSION.............................................................................................................33 4.1 Consolidation........................................................................................................33 4.1.1 Density Measurements...............................................................................33 4.1.2 Electron Microscopy Results......................................................................34 4.2 Solid Solution Formation......................................................................................34 4.3 Carbon Determination and Stoichiometry............................................................40 4.4 Comparison to Nitride Fuel Processing................................................................40 5 CONCLUSIONS AND FUTURE WORK.................................................................42 LIST OF REFERENCES...................................................................................................44 BIOGRAPHICAL SKETCH.............................................................................................46

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vii LIST OF TABLES Table page 3-1. Processing parameters and densities achieved..........................................................17 3-2. Stoichiometry of processed samples...........................................................................20

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viii LIST OF FIGURES Figure page 2-1. Glass reaction chamber with uran ium rod inside ready for hydriding.......................10 2-2. Uranium hydride powder flaking off of uranium metal rod.......................................10 2-3. Hydraulic press being loaded for cold pressing of green pellet................................11 2-4. Picture of spring-load ed hot press apparatus..............................................................13 3-1. Different magnification SEM, CU-1 with = 6.20 g/cc (55%TD)............................21 3-2. SEM with compositional contrast, CU-1 with = 6.20 g/cc (55%TD).....................22 3-3. SEM, CU-2 with = 6.52 g/cc (58% TD)..................................................................22 3-4. CU-7 with = 10.43 g/cc (92% TD), a)SEM b)SEM with compositional contrast...23 3-5. CU-8 with = 9.45 g/cc (84% TD), a)SEM b)SEM with compositional contrast.....24 3-6. CU-9 with = 9.08 g/cc (80% TD), a)SEM b)SEM with compositional contrast.....25 3-7. CU-10 with = 10.60 g/cc (94% TD), a)SEM b)SEM with compositional contrast.26 3-8. X-ray diffraction pattern for sample CU-1.................................................................27 3-9. X-ray diffraction pattern for sample CU-2.................................................................27 3-10. X-ray diffraction pattern for sample CU-3 or CU-4.................................................28 3-11. X-ray diffraction pattern for sample CU-5...............................................................28 3-12. X-ray diffraction pattern for sample CU-6...............................................................29 3-13. X-ray diffraction pattern for sample CU-7...............................................................29 3-14. X-ray diffraction pattern for sample CU-8...............................................................30 3-15. X-ray diffraction pattern for sample CU-9...............................................................30 3-16. X-ray diffraction pattern for sample CU-10.............................................................31

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ix 3-17. X-ray diffraction pattern for ZrC..............................................................................31 3-18. X-ray diffraction pattern for UC/UC2.......................................................................32 4-1. X-ray diffraction pattern of sample CU -1 compared with constituent carbides.........35 4-2. X-ray diffraction pattern of sample CU -2 compared with constituent carbides.........36 4-3. X-ray diffraction pattern of sample CU-3 /4 compared with constituent carbides.....36 4-4. X-ray diffraction pattern of sample CU -5 compared with constituent carbides.........37 4-5. X-ray diffraction pattern of sample CU -6 compared with constituent carbides.........37 4-6. X-ray diffraction pattern of sample CU -7 compared with constituent carbides.........38 4-7. X-ray diffraction pattern of sample CU -8 compared with constituent carbides.........38 4-8. X-ray diffraction pattern of sample CU -9 compared with constituent carbides.........39 4-9. X-ray diffraction pattern of sample CU -10 compared with constituent carbides.......39

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x Abstract of Thesis Presen ted to the Graduate School of the University of Florida in Partial Fulfillment of the Requirements for the Degree of Master of Science FEASIBILITY OF PROCESSING MIXED CARBIDE NUCLEAR FUELS FOR USE IN TRANSMUTATION SYSTEMS By Thomas Clifford Carter August 2005 Chair: Samim Anghaie Major Department: Nuclear and Radiological Engineering The Advanced Fuel Cycle Ini tiative (AFCI) is seeking a fuel form to be used as a target material for the transmutation of actinid es in either accelerator driven or reactor based fast spectrum systems. Currently funde d research by the AFCI fuel group has been in the areas of oxide, nitride, metallic, and dispersion fuels. This work focuses on the processing of mixed actinide/refractory metal car bides as a potential fuel to be used in these systems. The objective of this work was to produce a low porosity, solid solution, single phase, mixed carbide while staying with in parameters set by previous experience with processing minor actinides. Uranium was used as a surrogate for the actinides due to the constraint set by the available facilities. Ten samples of (U0.64,Zr0.36)C0.92 were processed by mixing powders, cold pressing at 180 and 300 MPa, and sintering for 10 hours at less than 2100 K in an induction furn ace. The upper limit of sintering was set by the hypothetical americium loss that would occur if it had actua lly been used in the study. Consolidation was difficult to achieve without pressing the samples at approximately 15

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xi MPa while sintering. The samples that were pressed during sinteri ng all showed a solid solution and single phase structure. A density of greater than 90% of the theoretical density was achieved in one sample. Previ ous experience has not shown that a solid solution could be formed at temperatures th at do not induce liquid pha se sintering. This work has shown that the application of a mode rate pressure during sintering helps for this solid solution formation to occur.

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1 CHAPTER 1 INTRODUCTION 1.1 Background Information The first commercial nuclear reactor in th e United States began operation in 1957. Nuclear power now constitutes about 20% of the electricity generated in the US and therefore produces a significant amount of hi ghly radioactive waste. Until a permanent repository is constructed and ope ned for operation, the waste is stored in pools of water, called spent fuel pools, at the site in which it was produced. It is the intention of the Department of Energy (DOE) to dispose of th is waste in a permanent repository at Yucca Mountain. It is estimated that the statutor y capacity of Yucca M ountain will be reached by the year 2015 and the theoretical capacity, if utilized, will be reached by the year 2050.1 When this limit is reached, there will be a need for a second repository if the current once-through fuel cycle continues to be the choice of the United States government. Spent nuclear fuel contains a large amount of useful elements that can be recycled in other reactors. Spent nucle ar fuel also contains some long-lived fission products that are responsible for a large amount of the heat load and radiotoxicity of the spent fuel. The isotopes that are useful for further power production after reproc essing include those of plutonium, neptunium, americium, and curium, the latter three of which are referred to as minor actinides. With the recycle of these isotopes, the radiotoxicity of the contents of a permanent repository is greatly reduced. An added benefit of reprocessing the plutonium in the spent fuel is that the repos itory will no longer contain it, which would be important in the future when the short-live d fission products have decayed away and the

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2 plutonium, a good material for use in nucle ar weapons, could be more easily mined creating a proliferation concern.1 1.1.1 Advanced Fuel Cycle Initiative Development Since 1957, when the first commercial reactor began operation, the US has operated on a once through fuel cycle. In re cent years, the DOE has recognized the need for a different solution to the problem of the disposal of spent fuel based on the previously presented facts. In 1999, DOE initiated the Accelerator Transmutation of Waste (ATW) program that carried the goal of utilizing high-powered proton accelerators for the transmutation of th e actinides in the fuel. In 2001, the ATW program was combined with the Accelerator Production of Tritium program to form the Advanced Accelerator Applications ( AAA) program. The major goal of AAA was to develop an accelerator-based test facility that would be used to transmute actin ides contained spent fuel. Again, in 2002, the goal of the program shifted to include re search into fast reactor concepts that would succeed in the ultimat e goal of transmuting minor actinides and possibly long-lived fission products. In recent years, this program has focused more and more on the use of fast reactor systems to achieve the transmutation. This change in focus prompted the DOE to change the name of the program to the Advanced Fuel Cycle Initiative (AFCI) in late 2002.2 1.1.2 Advanced Fuel Cycle Initiative Structure The AFCI is broken into four parts—separa tions, systems, materials, and fuels. Furthermore, the program is split into two pha ses of development. The first, Series I, focuses on the development of existing tec hnologies that will allow plutonium to be transmuted in current thermal spectrum light water reactors using a mixed oxide (MOX) fuel. This technology is expect ed to be deployable by the year 2010. Series II, which is a

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3 much longer term goal, expected to be deployable in the year 2030, focuses on the transmuting of minor actinides either through an accelerator driven system or a fast spectrum reactor. 1.1.2.1 Separations Chemical separations for Series I of th e AFCI program is focused on treatment of spent fuel from current reactors, which is in an oxide form. This involves breaking down the spent oxide fuel and sepa rating out particular fission pr oducts so that the plutonium mixed with uranium and be converted to a MOX fuel suitable for recycle in current operating nuclear reactors. Series II of the program will focus on treatment of spent fuel from Generation IV systems. These systems ha ve yet to be fully developed so these fuel types are not yet known; they could include carbides, oxides, nitrides, metals, or dispersion fuels. Once these fuels are broken down and the actinides are separated, they must be converted to a form suitable for us e as a feedstock in the manufacturing of the new fuel.3 1.1.2.2 Systems analysis The transmutation systems analysis group has been working in recent years to develop a suitable time frame and schedule of deployment. Also, this group is selecting which technology scheme will best suit th e goal of reducing the heat load and radiotoxicity of spent nuclear fuel, consideri ng all costs and benefits, while staying in line with the goals of the so-called Generation IV reactor program. A decision on what scheme and technologies are best suited for all goals of the DOE is expected in FY2007 to be given to the politicians w ho will make final decisions on the future of advanced fuel cycle research.4 1.1.2.3 Transmutation engineering The transmutation engineering group cove rs areas of materials development, reactor physics, and acce lerator technologies. Series I is focused on the development of

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4 existing technologies to burn plutonium and possibly other minor actinides in a mixed oxide fuel form. Series II focuses heavily on burning the higher actinides in either fast spectrum reactor systems or in accelerator dr iven systems. There are many technological advancements that must be made to make this goal feasible. This group works heavily on producing and verifying nuclear data and developing codes and models for the analysis of advanced fuel cycle systems. This is very important because these types of fast systems, which w ill be using actinides as fuel material, have not previously been considered in the area of traditional nuclear engineering. Also, structural material research is included in this category. This aspect of the program works very closely with research related to Generation IV system s that are currently being developed with DOE support. It is also a goal of the AFCI to develop equilibrium, high-burnup systems that will reduce the need for reprocessing the spent fuel.5 While the ADS is decreasingly being considered for use in transmuting actinides, this area of research is also included within the transmutation engineering. Projects to develop a lead-bismuth coolant for an ADS system are being completed as well as preliminary designs that couple a proton accelera tor with a subcritical assembly that will hold the actinides for transmutation.6 1.1.2.4 Fuels Research in the area of fuels for Series I is focused on the further development of mixed oxide fuels as well as determination of specifications of fuel that will be used in Generation IV systems for Series II. Deve lopment of MOX fuel for Series I closely follows previous research that was conducted in the area of MOX fuel for disposal of weapons grade plutonium.7

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5 Series II is focused on the development of different fuel forms that could be used for Generation IV systems in the long term. The fuel forms that are currently being considered are metal, nitride, oxide, and dispersion fuels.8 Ceramic fuel, including oxide, nitride, and carbide forms, offers an a dvantage of having high melting points while maintaining high burn-up potential This allows reactors to operate at high temperatures for long periods of time, which is a goal of the Generation IV program within DOE. Testing of these fuels is currently be ing conducted within the Advanced Test Reactor(ATR). Since this is a thermal reactor, these tests do not qualify as ideal for the purpose of developing a fuels for fast spectrum reactors; therefore, a fast flux booster is being planned for insertion into the ATR.9 1.1.3 History of Mixed Carbide Nuclear Fuel Long before the AFCI identified a need for a fuel to be used in the transmutation of minor actinides, mixed carbides had been studie d as a leading candidate for use in fast breeder reactors. The need for a high-b reeding ratio and lowplutonium containing reactor was pointed out by the Bethe Panel in 1973.10 This need began an effort to develop an advanced fuel system at a time when the feasibility of the fabrication of carbide fuels was in doubt. Irradiation pe rformance was also known to be poor due to swelling behavior, fission gas release, and stru ctural changes. Sin ce this time, many of these problems have been addressed and solved.11 The Experimental Breeder Reactor-II cont ained test fuel assemblies that used (U0.24, Pu0.76)C as fuel. These assemblies, wh en sodium bonded, achieved burnups of 15.8 at% before failure; however, the helium bonded pins achieved burnups of 20.7 at% without failure. Both of these test s used 316 Stainless Steel as cladding.12 The Fast Breeder Test Reactor, in India, also used a mixed uranium-plutonium fuel. In this

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6 reactor, the fuel material was (U0.3, Pu0.7)C, chosen so that un-enriched uranium could be used.13 There have also been a numb er of High Temperature Gas-Cooled reactors that have used carbide fuels, including Peach Bottom Nu clear Power Station and Fort St. Vrain in the United States.13 Both reactors mentioned used gr aphite coated microspheres of (U, Th)C2 or UC2 embedded in a graphite matrix. These reactors were intended to achieve high burnup in the thermal neutron spectrum. Fort St. Vrain was expected to reach an average burnup of 100 MWD/kg before it was shut down due to economic influences.14 1.2 Purpose of Work 1.2.1 Motivation The motivation behind this project lies in producing a fuel that will contain actinides as a target for the transmutation a nd destruction of long liv ed actinides such as Np-237 and Am-241, which is a goal of the AFCI. The desired characteri stics of this fuel include a form that can achieve high burn-up, operate and be chemically stable at high temperatures, and contain the fission products. It has been indicat ed that major funding for AFCI Fuels has not traditionally incl uded carbides. Mixed carbides not only offer advantages that all ceramics do, but ther e are also significant advantages in the processing of these fuels. 1.2.2 Objective The objective of this project is to fabricat e low porosity, solid solution, single phase mixed carbides that will provide the desi red characteristics of the AFCI using an appropriate surrogate for the actinides of concern. Als o, it is important that the development of the processing technique a nd procedure take into consideration the volatility of the actinides of concern, namely americium. Also, a comparison of the carbides produced in this st udy with the nitrides produced by others will be presented

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7 since nitride fuel shows the most similar ope rating characteristics of the fuels being considered by the AFCI.

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8 CHAPTER 2 METHODS AND MATERIALS 2.1 Processing 2.1.1 Powder Processing 2.1.1.1 Composition The composition of the samples processed was chosen to be (U0.64, Zr0.36)C0.92. This corresponds with work completed by Margevicius15 in the area of nitride fuel. This work uses uranium as a surrogate for all actinides (plutonium, americium, neptunium, and cerium). The purpose of using this surr ogate is for the ease of working with available materials within the available facili ties. The use of remote handling systems for highly radioactive actinides is beyond of the scope of this study. Also, a hypostoichiometric composition was chosen to prevent the formation of a second phase, carbon, which would substantially lower the melting point of the fuel. A total of 10 samples were prepared using the above composition. 2.1.1.2 Powder mixing The above composition was achieved by mi xing appropriate amounts of ZrC, UH3, graphite, and binder. The ZrC was obtained from Alfa Aesar, Lot# F10E09 and has a purity of 98%. The graphite was obtained from Johnson Mathey, Lot# I24C08 and has a purity of 99.5%. Stearic acid was used as a bi nder and included in the mixture at 3% of the total mass. This stearic acid was obtained from Aldrich, Lot# 07112AF and has a purity of 95%. The uranium hydride was pr oduced using natural uranium rod and the description of this process is described in the following section.

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9 The powders and binder were weighed using a Sartorius model R180D balance and placed in a 125 mL Nalgene HDPE bottle. Approximately 20 chromium balls (diameter 0.635 cm) were placed in the bottle with the powder before the bottle was placed on a Lortone model 1.5E rotary tumbler for mixi ng. Mixing was completed overnight for at least 18 hrs. The handling of these powders was done primarily in a fume hood with the exception of adding the uranium hydride, which is pyrophoric in air. This handling was done in a glove box under an inert helium atmo sphere. In the glove box, an AND model HL-200 balance was used for the purpose of weighing the uranium hydride. Uranium hydride processing. Uranium hydride was produced using uranium metal rodlets of natural enrichment while flow ing an Ar-7%H mixture over it at 473 K. Heating was accomplished using a Thermodyne Type 1900 Hot Plate. This reaction took place in a custom made reaction vessel, whic h is shown in Figure 2-1, contained in a fume hood. As UH3 is formed by the hydrogen reacting with the uranium, it flakes off leaving more uranium metal exposed, as s hown in figure 2-2. This process took approximately 48 hrs to complete for a piece of uranium weighing 7g. Mixing this UH3 with the other constituent powders allo wed for the desired hypo-stoichiometric composition to be obtained. During the early phase of sintering, at approximately 676 K, the hydrogen is evolved through the decomposition of the UH3.15 2.1.2 Cold Uniaxial Pressing Once the powders were sufficiently mixed, a 2 g sample was taken from the bottle and placed into a stainless stee l die. This die has a diam eter of 0.635 cm with a small plug inserted in the bottom and a longer punch th at is inserted from the top to allow for pressing. Pressures of 180 and 300 MPa were used, each for five of the samples. The actual pressing of the sample was accomp lished using a Carver model #3912 hydraulic

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10 jack attached to a steel block. The top of the punch was in constant contact with a permanently mounted steel block above. Th is apparatus allowed for pressing well above the needed pressure. A picture of th e hydraulic press is shown in Figure 2-3. Figure 2-1. Glass reaction chamber with ur anium rod inside ready for hydriding. Figure 2-2. Uranium hydride powder fl aking off of uranium metal rod.

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11 Figure 2-3. Hydraulic press being loaded for cold pressing of green pellet. 2.1.3 Sintering and Hot Press Following cold pressing, the sample was placed into a graphite susceptor for sintering. A Taylor Winfield, mode l CE2000, 20 kW, 450 kHz induction furnace was used for sintering. Two Insulator Seal, In c., model 9511020, water cooled electrical feedthroughs, rated for 10 kV, 35 kW were welded into two separate 2 in. flanges that were attached to a stainless-steel water-cooled te st chamber, containing a helium atmosphere, with a starting pressure of 250 torr. These we re attached to a 4-turn 3.3 cm ID coil made from thick-walled (0.124 cm) copper tubing. A 1.6 cm diameter, 8 cm long graphite susceptor, with a 0.635 cm hole drilled for sa mple containment was used. The sample would sit in the graphite suscep tor on a graphite punch that wa s in contact with a graphite block below. Samples CU-1 through CU-6 were sintered with no pressure being applied to the sample inside of the susceptor. These samples, as will be seen in future chapters, showed

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12 very poor consolidation. Ther efore, a spring-loaded hot pr ess apparatus was designed to sit inside of the test cham ber to apply a pressure of approximately 15 MPa during sintering. This pressure was de pendent on the length of the samp le due to the fact that it was spring-loaded. The hot-press apparatus is shown in Figure 2-4 with a susceptor loaded on a graphite stand. The press is ma de of stainless steel and sits on a piece of mica, to prevent electrical coupl ing with the test chamber. The graphite block that the susceptor rests on is used to assure that the sample sits in the middl e of the coil axially. The test piece is completely electrically is olated from the test piece by a alumina tube present above the sample and a short length of mullite pipe underneath the graphite block on which the susceptor rests. Unfortunately, because of the configuration of the chamber, it was impossible to measure the compaction of the sample during sint ering. This presents a slight problem in the confidence of the pressure that is applied throughout th e entire sintering process due to the change in the length of the sample during compaction. 2.1.3.1 Temperature measurement and control Temperature was monitored and contro lled using a Maxline™ Temperature Acquisition and Control System. This used tw o, “two-color” infrared thermometers also made by Maxline, model #MX-MR04, which have a temperature range of 977 to 3866 K. These sensors measure the ratio of energy emitted at two separate wavelengths, 0.7 and 1.7 microns. Error is introduced in this measurement by the wavelength dependence of the emissivity of the target being measure d. This is accounted for in the system by setting the emissivity slope ( slope), which is determined by the manufacturer for different materials. For graphite no correction is needed so the slope is set to 1.0.

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13 Figure 2-4. Picture of spring-lo aded hot press apparatus. 2.1.3.2 Sintering schedule To parallel the work done in nitride fuel development, samples were processed at temperatures between 1900 and 2100 K. It was a goal of this study to produce high density (low porosity) carbide fuel at proces sing temperatures lower than about 2100 K to prevent the hypothetical americium loss that would occur if the samples actually contained actinides other th an uranium as would be th e case for fuel used in transmutation systems. Samples were processed for times of 10 and 5 hrs. The work in nitride fuel only processed samples for 10 hr s, however this study included the shorter sintering time for the purpose of determining whether the long sintering time added a benefit to the density of the sample. Early samples, that had no pressure applied in the

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14 furnace were processed two at a time with a gr aphite plug in between them to prevent any diffusion bonding of the samples. 2.2 Characterization 2.2.1 Density Measurements In order to provide a measure of porosit y, the bulk density of the samples was measured. Also, the measurement of density gives an indication of the degree of sintering and provides information on whet her or not additional processing will be needed. Because of the somewhat irregular shapes of the samples the volume was measured by determining how much water wa s displaced from a known amount. First, the dry samples were weighed. Then a small container of water, approximately 1 cm3 was weighed before placing the sample in the small container along with the water and weighed once again. The volume of the sample was calculated by taking the sum of the masses of the sample and the container with only water and subtracting the mass of sample and water together. Since the dens ity of water is 1 g/cc, the mass difference calculated gives the volume of the sample. There was some error in this measurement because of the difficulty in assuring that th e meniscus in the container was exactly the same size each time a measurement was take n as well as the irregularity in shapes holding small pockets of air under the water. Therefore, five measurements were taken and averaged together. The density was compared to an ideal solid solution calculation of the constituents. This calculation used lattice parameters that were a weighted av erage of the lattice parameters of the constituent carbides as expected from VegardÂ’s Law.

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15 2.2.2 Scanning Electron Microscopy Scanning Electron Microscopy (SEM) was performed on all samples that were consolidated enough for polishing. For preparation, samples were ground using traditional metallurgical techniques and polis hed with diamond paste down to 10 micron grit size. Micrographs were taken using a JEOL JSM 6400 with a Link ISIS digital image capture system. All micrographs we re taken at a 15 mm working distance and a 15 kV accelerating voltage. 2.2.3 X-Ray Diffraction Slides were prepared for X-Ray Diffr action (XRD) by pulverizing small pieces of the sample into a powder. These were pla ced on a slide and a solution of one part collodian and seven parts amyl acetate was used for adhesion purposes. A Philips APD 3720, operating at 40kV, with a current of 20 mA was used for the actual measurement of the samples. Results of were compared to known XRD spectra from the constituent carbides to assure solid solution characteristics. 2.2.4 Carbon Determination Another small portion (about 0.25 g) of the samples was taken and pulverized with a mortar and pestle for C/M ratio anal ysis. After these we re sufficiently ground, they were placed in a LECO WC-200 Car bon Determinator. This apparatus uses induction heating to combust the sample and measure the amount of carbon dioxide and carbon monoxide produced. This information gives a weight percent (based on input mass) of carbon in the sample.

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16 CHAPTER 3 RESULTS 3.1 Consolidation and Density In all, ten samples of (U0.64, Zr0.36)C0.92 were fabricated. Of these samples, six showed consolidation sufficient for characteri zation using the techniqu es described in the previous chapter. Table 3-1 shows the pro cessing parameters and resultant density of each of the samples. As shown, samples CU -3, CU-4, CU-5, and CU-6 did not achieve sufficient consolidation for measurement. Furthermore, sample CU-4 was completely blown apart during sintering a nd was not contained inside of the graphite susceptor. Because of this, the heating rate for subsequent samples was lowered substantially in order to prevent hydrogen from the decomposition of UH3 and the off-gassing of the binder from causing destruction of the pellet. Samples CU-5 and CU-6 appeared to have sufficient consolidation; however, during gr inding and polishing, the samples did not hold up and were ground down quite quickly. It should be noted that the uncertainty of measurement of sample CU-7 is very high which causes the measurement to be suspect. This was due to the fact that density measurements were performed after parts of th e samples were taken off for other testing. For sample CU-7 this left a very small am ount for measurement and, since the technique employed for this measurement works best for samples near the control volume of water, a larger amount of uncertainty was introduced into this measurement.

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17 Table 3-1. Processing paramete rs and densities achieved. Not consolidated enough for measurement. 3.2 Scanning Electron Microscopy Analysis Figures 3-1 through 3-7, found in a portfolio at the end of this chapter, show SEM micrographs of only the consolidated samp les after polishing – sa mples CU-3 through CU-6 were not analyzed with SEM because they were not consolidated enough to be sufficiently polished. All samples have sec ondary electron images which show general topographical characteristics. Where possible, a second micrograph of the samples was taken using backscattered electrons to show compositional contrast. Lighter areas in these images indicate a higher average atomic number than the darker areas. Because the atomic number of Zr (Z=40) is lower than that of U (Z=92) it is noted that the lighter areas in the images represent higher concen trations of uranium in the microstructure. Micrographs for CU-1 and CU-2 show that these samples are barely consolidated and the porosity is extremely high (55% TD and 58% TD, respectively). Figure 3-2 shows an SEM with compositional contrast of sample CU-1. It is noted that there is a large amount of lighter areas wh ich indicate that there is so me residual uranium carbide Sample Pre-Sinter Pressure (MPa) 15 MPa Press During Sintering Sinter Temperature (K) Sinter Time (hrs) Average Density (g/cc) % Theoretical Density CU-1 180 No 1923 10 6.20 55 CU-2 300 No 1923 10 6.52 58 CU-3 180 No 1923 5 * CU-4 300 No 1923 5 * CU-5 180 No 2023 10 * CU-6 300 No 2023 10 * CU-7 180 Yes 1973 10 10.43 92 CU-8 300 Yes 1973 10 9.45 84 CU-9 180 Yes 2023 10 9.08 80 CU-10 300 Yes 2023 10 10.60 94

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18 or uranium that has not sufficiently diffused and formed a solid solution. This will be further evidenced by the x-ray diffraction result s presented later. Sample CU-2 is shown in Figure 3-3. It shows that it has achieved a similar level of consolidation as sample CU-1 with a large amount of porosity. Figure 3-4 shows a much smaller amount of porosity than the previous figures which indicates a much higher level of sinter ing for sample CU-7. Also the rounded and elongated pores indicate an intermediate leve l of sintering. Also, from the SEM with compositional contrast, it is seen that ther e are small areas of higher uranium content which would indicate some larger scale inhom ogeneity. It should also be noted that the higher concentration of uranium carbide is not found in the grain boundaries or porous regions which indicates that these samples did not achieve any liquid phase sintering which is expected since the processing temp eratures were not close to the melting point of uranium carbide (3063 K). Figure 3-5 shows areas of sample CU-8 that have high levels of consolidation with low porosity as well as areas that have a hi gher porosity. Again, like sample CU-7, the micrograph shows areas of higher uraniu m concentration which indicates some inhomogeneity. This sample also shows rounded and elongated pores indicating an intermediate level of sintering. Figures 3-6 shows a consolidation and por osity consistent with the measured density for sample CU-9. There is very l ittle contrast in the SEM with compositional contrast which indicates a homogeneous sample. Figure 3-7 shows sample CU-10 and doe s not correspond well to the measured density. This is believed to be the result of the SEM micrographs to taken well after all of the samples were produced. This may ha ve allowed some oxidation of the areas with

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19 higher uranium concentrations to occur. After th is occurred, it is likely that some parts of the sample lifted out during polishing and th erefore left what appears to be pores. 3.3 X-Ray Diffraction Figures 3-8 through 3-16 show X-ray diffrac tion patterns for each of the fabricated samples. As expected from earlier results, samples CU-1 through CU-6 do not appear to be solid-solution. The latter diffraction pa tterns of samples CU-7 through CU-10 all indicate that the samples achieved a solid solution. Samples CU-3 and CU-4 are presented together since only a small fragme nt of sample along with residual loose powders was found in the susceptor, which caused the two samples to be indistinguishable. Each sample will be discussed further in Chapter Four. Figure 3-17 shows the x-ray diffraction patte rn for the starting zirconium carbide powder and Figure 3-18 shows a diffraction pa ttern for uranium carbide powder. These will be used in the next chapter for comparison of the processed samples with the individual constituents. It should be noted that the uranium carbide powder is a mixture of UC and UC2; this is the composition that was r eceived from the vendor. Although the uranium carbide powder was not used to make the powder for processing, the uranium, in the form of UH3, quickly reacts with the free car bon to form uranium carbide once the UH3 has dissociated. 3.4 Carbon Determination All samples were analyzed by the LECO WC-200 Carbon Analyz er in order to determine the exact stoichiometry of each samp le. These results are presented in Table 3-2. The stoichiometry shown in the table indi cates what it would be if the samples were, in fact, solid solution and single phase. This is obviously not true for those samples that did not achieve a solid solution, single-phase structure, which will be discussed in future

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20 sections. Again, samples CU-3 and CU-4 ar e presented together for the same reason mentioned above. Further discussion of thes e results will be presented in Chapter Four. Table 3-2. Stoichiometry of processed samples. Sample As Mixed Stochiometry Final Carbon Weight Percent Final Stoichiometry CU-1 (U0.64, Zr0.36)C0.92 8.88 (U0.64, Zr0.36)C1.5 CU-2 (U0.64, Zr0.36)C0.92 6.25 (U0.64, Zr0.36)C1.02 CU-3/CU-4 (U0.64, Zr0.36)C0.92 5.10 (U0.64, Zr0.36)C0.83 CU-5 (U0.64, Zr0.36)C0.92 3.47 (U0.64, Zr0.36)C0.55 CU-6 (U0.64, Zr0.36)C0.92 4.22 (U0.64, Zr0.36)C0.68 CU-7 (U0.64, Zr0.36)C0.92 4.47 (U0.64, Zr0.36)C0.72 CU-8 (U0.64, Zr0.36)C0.92 4.55 (U0.64, Zr0.36)C0.74 CU-9 (U0.64, Zr0.36)C0.92 4.28 (U0.64, Zr0.36)C0.69 CU-10 (U0.64, Zr0.36)C0.92 3.81 (U0.64, Zr0.36)C0.61

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21 a. b. Figure 3-1. Different magnifi cation SEM, CU-1 with = 6.20 g/cc (55%TD)

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22 Figure 3-2. SEM with compositi onal contrast, CU-1 with = 6.20 g/cc (55%TD) Figure 3-3. SEM, CU-2 with = 6.52 g/cc (58% TD)

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23 a. b. Figure 3-4. CU-7 with = 10.43 g/cc (92% TD), a)SEM b)SEM with compositional contrast.

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24 a. b. Figure 3-5. CU-8 with = 9.45 g/cc (84% TD), a)SEM b)SEM with compositional contrast.

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25 a. b. Figure 3-6. CU-9 with = 9.08 g/cc (80% TD), a)SEM b)SEM with compositional contrast.

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26 a. b. Figure 3-7. CU-10 with = 10.60 g/cc (94% TD), a)SEM b)SEM with compositional contrast.

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27 0 50 100 150 200 250 300 350 1030507090 2 (degrees)counts Figure 3-8. X-ray diffraction pattern for sample CU-1. 0 100 200 300 400 500 600 1030507090 2 (degrees)counts Figure 3-9. X-ray diffraction pattern for sample CU-2.

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28 0 100 200 300 400 500 600 700 800 1030507090 2 (degrees)counts Figure 3-10. X-ray diffraction patte rn for sample CU-3 or CU-4. 0 100 200 300 400 500 600 700 800 1030507090 2 (degrees)counts Figure 3-11. X-ray diffraction pattern for sample CU-5.

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29 0 200 400 600 800 1000 1200 1400 1030507090 2 (degrees)counts Figure 3-12. X-ray diffraction pattern for sample CU-6 0 50 100 150 200 250 300 350 400 450 500 1030507090 2 (degrees)counts Figure 3-13. X-ray diffraction pattern for sample CU-7

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30 0 100 200 300 400 500 600 1030507090 2 (degrees)counts Figure 3-14. X-ray diffraction pattern for sample CU-8 0 100 200 300 400 500 600 700 1030507090 2 (degrees)counts Figure 3-15. X-ray diffraction pattern for sample CU-9.

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31 0 100 200 300 400 500 600 700 1030507090 2 (degrees)counts Figure 3-16. X-ray diffraction pattern for sample CU-10. 0 1000 2000 3000 4000 5000 6000 1030507090 2 (degrees)counts Figure 3-17. X-ray diffrac tion pattern for ZrC.

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32 0 50 100 150 200 250 300 350 400 20406080100 2 (degrees)counts Figure 3-18. X-ray diffrac tion pattern for UC/UC2.

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33 CHAPTER 4 DISCUSSION Each of the characterization methods de scribed in Chapter 2 and the results presented in Chapter 3 will be used as a t ool to asses whether the samples produced in this study accomplish the objectives as point ed out in Chapter 1. Consolidation and density are a measure of porosity. The de nsity is measured both by the direct measurement and analysis of the SEM mi crographs. Solid solution formation is measured by the analysis of the XRD spectra to examine the crystal structure. The phase structure of the samples was assessed by the analysis of the XRD spectra and the C/M ratio based on results from the LECO Carbon Determinator. 4.1 Consolidation 4.1.1 Density Measurements The results presented in the previous ch apter have shown that consolidation of samples, into anything more than a slightly bound powder, cannot occu r at the prescribed temperatures without some hot pressing be ing involved during si ntering. This is evidenced by the fact that most of the sa mples without pressing did not produce a solid pellet at all and those that di d were difficult or impossible to analyze with SEM because of their weak mechanical structure that did not allow for polishing. The samples with the highest density were produced by sintering at temperatures at or above 1973 K for 10 hours. It is also noted that th e magnitude of the pre-sinter press seemed to have an effect on the final density of the pellets. This eff ect is shown by each pair of pellets that were sintered simultaneously after being pressed with pressure s of 180 and 300 MPa prior to sintering. The higher cold-press pressure s resulted in higher bulk densities.

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34 4.1.2 Electron Microscopy Results Microscopy results have proven to fo llow the trend of the actual density measurements in most cases. Two of the samples, however, showed slight deviation from the measured results. The SEM micros copy for Sample 7 does not confirm the high (92% TD) measured density. This deviat ion is accounted for by the high level of uncertainty that was found in the measurement of this density. Microscopy for Sample 10 shows a slight devi ation from what is expected since the density measurement showed this sample to be 94% TD and the SEM results show a much more porous sample. This is likely to be caused by a delay in actual time between the density measurement of this sample and the microscopy preparation of the sample. A long period of time between these measuremen ts could have caused some areas become oxidized due to the fact that the sample was not stored in an inert atmosphere. This oxidation causes these areas to be uninte ntionally removed during the grinding and polishing of the sample. 4.2 Solid Solution Formation As shown in Figure 4-1, sample CU-1, which was sintered at 1923 K for 10 hours with no hot-pressing, shows a large amount of free zirconium carbide in the sample, which indicates that it did not form a solid so lution. Also, there is indication of a second UC2 phase formed in this sample, which w ould agree with the results of the carbon determination that indicated a hyperstoichiometric composition. Samples CU-2 through CU-6, as indicated by Figures 4-2 through 4-5, all show patterns that reflect a nearly complete solid solution formation with some concentration gradients, which is indicated by the broadening of th e peaks. It should be noted that, in these patterns, there are very small peaks in line with the zirconium carbide peaks, indicating a small amount of zi rconium carbide left in the sa mple. These results are in

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35 agreement with the SEM micrographs that sh ow lighter and darker areas that would indicate higher areas of uranium and zirconium, respectively. Samples CU-7 through CU-10, as indicat ed in Figures 4-6 through 4-9 all show very sharp peaks corresponding with d-values consistent with a mixed solid solution mono-carbide formation. There is no indicati on of other phases present in the sample and no indication of residual star ting powders. The peaks for the samples are located between those of the zirconium carbide a nd uranium carbide, in accordance with the composition that was mixed, (U0.64,Zr0.36)C0.92. 0 0.2 0.4 0.6 0.8 1 1.2 20406080100 2 (degrees)Relative Intensity CU-1 ZrC UC/UC2 Figure 4-1. X-ray diffraction pattern of sample CU-1 compared with constituent carbides.

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36 0 0.2 0.4 0.6 0.8 1 1.2 20406080100 2 (degrees)Relative Intensity CU-2 ZrC UC/UC2 Figure 4-2. X-ray diffraction pattern of sample CU-2 compared with constituent carbides. 0 0.2 0.4 0.6 0.8 1 1.2 20406080100 2 (degrees)Relative Intensity CU-3/4 ZrC UC/UC2 Figure 4-3. X-ray diffraction pattern of sa mple CU-3/4 compared with constituent carbides.

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37 0 0.2 0.4 0.6 0.8 1 1.2 20406080100 2 (degrees)Relative Intensity CU-5 ZrC UC/UC2 Figure 4-4. X-ray diffraction pattern of sample CU-5 compared with constituent carbides. 0 0.2 0.4 0.6 0.8 1 1.2 20406080100 2 (degrees)Reative Intensity CU-6 ZrC UC/UC2 Figure 4-5. X-ray diffraction pattern of sample CU-6 compared with constituent carbides.

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38 0 0.2 0.4 0.6 0.8 1 1.2 20406080100 2 (degrees)Relative Intensity CU-7 ZrC UC/UC2 Figure 4-6. X-ray diffraction pattern of sample CU-7 compared with constituent carbides. 0 0.2 0.4 0.6 0.8 1 1.2 20406080100 2 (degrees)Relative Intensity CU-8 ZrC UC/UC2 Figure 4-7. X-ray diffraction pattern of sample CU-8 compared with constituent carbides.

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39 0 0.2 0.4 0.6 0.8 1 1.2 20406080100 2 (degrees)Relative Intensity CU-9 ZrC UC/UC2 Figure 4-8. X-ray diffraction pattern of sample CU-9 compared with constituent carbides. 0 0.2 0.4 0.6 0.8 1 1.2 20406080100 2 (degrees)Relative Intensity CU-10 ZrC UC/UC2 Figure 4-9. X-ray diffraction pattern of sample CU-10 compared with constituent carbides.

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40 4.3 Carbon Determination and Stoichiometry The results of the carbon determination, Ta ble 3-2 showed that samples CU-1 and CU-2 had a significant increase in the carbon co ntent. This is also indicated by the formation of a UC2 phase which is shown to be pres ent in CU-1 by the x-ray diffraction results. This increase in carbon content is at tributed to “carbon pi ck-up” as described by Knight16. The results for the remaining samples, howev er, seem to be counter intuitive. They show that the carbon content has decreased sign ificantly in these samples. It should be noted that the LECO Carbon Determinator was calibrated with a tungsten carbide sample just before analysis was performed on the fabr icated samples. This decrease in carbon content remains unexplained because there is no reason for the sample to lose carbon into a graphite susceptor, where the sample was pr ocessed. It should be further noted, that although the lower carbon content and stoich iometry would indicate that the sample would take on a lower phase of carbide, i.e. sesquicarbide-M2C3, there is no indication of any other phase than the expected monocarbide-MC in the XRD results. 4.4 Comparison to Nitride Fuel Processing All attempts were made to parallel early work completed by Margevicius in composition and processing parameters to complete a fair comparison of results of the processing of carbide to nitride fuels. Margev icius et. al.’s work showed that the highest density achieved in processing the nitride fuels was 88% TD.17 This work, namely in sample CU-10, achieved a density of >90% TD following the same sintering schedule. It should again be noted that, although the ratio of actinide to zirconium has remained the same between the studies, this study has used ur anium as a surrogate for all actinides. It is expected that as higher actinides are in troduced into the system, higher densities may

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41 be achieved due to the lower melting point of the other actinide carbides, such as plutonium carbide, americium carbide, etc. Unfortunately, there has been no data published of x-ray diffraction results indicating the structure of the final nitride fuel pellets. It is indicated, however, in a presentation by Margevicius17 that original irradiation testing of the first fabricated fuel pellets showed that they were rather brit tle and could not hold up mechanically to the helium and fission gas release. The fabrication process had to be modified, using slightly higher temperatures, to “solut ionize” the powders before the carbothermic reaction, cold pressing, and sintering.

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42 CHAPTER 5 CONCLUSIONS AND FUTURE WORK The main objective of this work was to develop a low porosity (>90% TD), solid solution, single phase fuel that could potentially be used in AFCI related fast spectrum transmutation systems. This objective wa s essentially met in sample CU-10. This sample was processed as described above by mixing the constituent powders at the designated composition, cold-pressing at 300 MPa, and sintering at 2073 K under a moderate pressure of 15 MPa. The sintering conditions used for this sample do stay within the prescribed temperature range (les s than 2100K) as designated by the volatility of americium that will eventually be the actinide used in the fuel. This sampleÂ’s properties, at least, ma tch the best nitride samples produced by Margevicius et. al. Carbide fuels offer signif icant advantages over ni tride fuels in that there are no radioactive daughters of na tural carbon isotopes as there is for 14N (14N(n,p)14C), the primary constituent of naturally occurring nitrogen. Because there are no radioactive daughters of natural carbon, there is no need for any enrichment process. Also, the conversion of the actinide oxide feed stock into actinide carbide does not require any partial pressure of nitrogen gas to be held during that ste p, if a uranium oxide feedstock is chosen. These advantages indicat e that mixed carbides should be considered as a highly advantageous potential fuel form for the Advanced Fuel Cycle Initiative. It is also of interest, although not a speci fic objective of this study, that the binary carbide samples that were processed using a mo derate pressure during sintering appear to have achieved a solid solution. This is signif icant because it is not evident from previous

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43 literature that solid solution fo rmation has been achieved in the past without sintering at much higher temperatures to induce liquid-ph ase sintering. Operat ing at these extreme temperatures, often above 2800 K (for mixed ur anium carbides) has proven to be very difficult. The XRD results in this study clearly show solid solution formation while sintering at only moderate temp eratures of 2023 K, but only when a hot press was used to introduce approximately 15 MPa uniaxially during sintering. There are many aspects of this study that would be ideal for continuing research. Unless facilities and materials are available for the fabrication of the actual compositions desired, the next step in this process would be to fabricat e a variety of samples using surrogates for each individual actinide consider ed. Surrogate selection should be based on non-radioactive lanthanides that have si milar melting points and vapor pressures. Close attention would need to be paid to th e loss of individual elem ents in these samples indicating what the loss might be for the re spective actinides. This would give a good understanding of the process i nvolved in producing these carbide fuels without the use of expensive materials and facilities that would be involved with handling the actinides of interest.

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44 LIST OF REFERENCES 1. J. W. HERCZEG, Program Overview: Sp ent Fuel Treatment, Reprocessing, and Transmutation, Presented to Advanced Fu el Cycle Initiative Fellows, Chicago, IL (July 2002). 2. J. W. HERCZEG, Advanced Fuel Cycle In itiative Semi Annual Review: Opening Remarks, Presented at AFCI Semi-Annual Review Meeting, Albuquerque, NM (January 2003) 3. J. LAIDLER, Development of Separations Technologies under the Advanced Fuel Cycle Initiative, Presented at AFCI Semi-Annual Review Meeting, Albuquerque, NM (January 2003) 4. R. G. BENNETT, AFCI Systems Analysis Overview, Presented at AFCI Semi-Annual Review Meeting, Santa Fe, NM (August 2003) 5. M. W. CAPIELLO, Transmutation Engineering: Introduction, Presented at AFCI Semi-Annual Review Meeting, Albuquerque, NM (January 2003) 6. M. W. CAPIELLO, AFCI Transmut ation Engineering Overview, Presented at AFCI Semi-Annual Review Meeting, Santa Fe, NM (August 2003) 7. K. O. PASAMEHMETOGLU, AFCI Fuels Development Overview, Presented at AFCI Semi-Annual Review Meeting, Santa Fe, NM (August 2003) 8. K. O. PASAMEHMETOGLU, AFCI Fuels Development Overview," Presented at AFCI Semi-Annual Review Meeting, Albuquerque, NM (January 2003) 9. J. M. RYSKAMP, Fast Neutron Flux Booster in the Advanced Test Reactor, Presented at SFCI Semi-Annual Review Meeting, Albuquerque, NM (January 2003) 10. H.A.BETHE, Advanced Nuclear Power, Report of the Cornell Workshops on the Major Issues of a National Energy Res earch and Development Program, Cornell University, Ithaca, NY, 169 (1973) 11. R. B. MATTHEWS and R. J. HERBST, Uranium-Plutonium Carbide Fuel for Fast Breeder Reactors, Nuclear Technology 63, (1983) 9 12. G. R. HARRY, Status of Steady-State I rradiation Testing of Mixed-Carbide Fuel Designs, LA-UR-83-1248, Los Alam os National Laboratory (1983)

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45 13. H. J. MATZKE, Science of Advanced LMFBR Fuels Elsevier Science Publishers, Amsterdam, The Netherlands (1986) 14. M. M. EL-WAKIL, Nuclear Energy Conversion American Nuclear Society, La Grange Park, IL (1982) 15. R. W. MARGEVICIUS, AFCI Fuels Development Update, Presented at AFCI SemiAnnual Review Meeting, Albuquerque, NM (2003) 16. T. W. KNIGHT, Processing of Solid So lution Mixed Uranium/Refractory Metal Carbides for Advanced Space Nuclea r Power and Propulsion Systems, PhD Dissertation, University of Florida, Gainesville (2000) 17. R.W. MARGEVICIUS, Nitride Fuel Fabr ication Efforts at Los Alamos for Advanced Fuel Cycle Initiative, Presented at AFCI Semi-Annual Review Meeting, Santa Fe, NM (2003)

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46 BIOGRAPHICAL SKETCH Thomas C. Carter was born in Vacav ille, CA, on February 2, 1981. He moved around the country in the early y ears of his life, living in Ca lifornia, Ohio, and Georgia, before moving to Fort Walton Beach, FL, in January of 1995. He graduated magna cum laude from Fort Walton Beach High School in June 1999. From there, Thomas attended the University of Florida and majored in nuc lear engineering. He graduated from the Department of Nuclear and Radiological Engineering with high honors in May 2002. After participating in an in ternship at Argonne National La boratory in the Chemical Technology Division he returned to the Univer sity of Florida to pursue his Master of Science in nuclear engineering sciences. Du ring this time, he was a recipient of the Advanced Fuel Cycle Initiative Graduate Fellowship.