Symposium Organizers
Ram Devanathan Pacific Northwest National Laboratory
Robin W. Grimes Imperial College London
Kazuhiro Yasuda Kyushu University
Blas Pedro Uberuaga Los Alamos National Laboratory
Constantin Meis CEA, INSTN-UESMS
T1: Structural Materials I
Session Chairs
Monday PM, November 26, 2007
Gardner (Sheraton)
10:00 AM - **T1.1
From Bytes to Ingots: Design of Structural Materials for Advanced Nuclear Energy Systems.
Steven Zinkle 1 , Ronald Klueh 1 , Philip Maziasz 1 , Jeremy Busby 1 , Lance Snead 1 , Roger Stoller 1 , John Vitek 1 , Roger Stoller 1 , Yury Osetsky 1
1 Materials Science & Technology Division, Oak Ridge National Lab, Oak Ridge, Tennessee, United States
Show Abstract10:30 AM - T1.2
Microstructural Evolution in Esshete 1250 Austenitic Creep Resistant Steel.
Chia Wong 1 , Hardy Mohrbacher 2 , Mike Spindler 3 , Klaus Hulka 2 , George Fourlaris 1
1 School of Engineering, Swansea University, Swansea United Kingdom, 2 , Niobium Products GmbH, Dusseldorf Germany, 3 Assessment Technology Group, British Energy, Gloucester United Kingdom
Show AbstractAustenitic steels are primary candidates for superheater or reheater tubing, where oxidation resistance and fireside corrosion become important in addition to creep strength required in power station applications. Esshete 1250 is an austenitic CrNiMn steel with high strength relative to other ordinary austenitic stainless steels used in power generation. This grade is designed with the properties of high ductility and high strength at elevated temperatures up to 675°C, combined with a desired microstructure and mechanical properties which remain stable over a long period, easily fabricated and welded using standard procedures as well as good oxidation resistance at elevated temperature application. Esshete 1250 has been widely used and is a choice of materials for superheater boiler tube in UK power stations. An experimental study of the microstructure evolution under severe creep exposure conditions has been carried out on Esshete 1250 austenitic stainless steel. The microstructural evolution and its relationship to creep rupture in Esshete 1250 Niobium alloyed austenitic creep resistant alloy will be outlined in this paper. It is well known that in pure metals or single phase solid solution alloys, grain size markedly increase the strength thus creep resistance. Factors that could influence the creep resistance is the formation, dissolution and coarsening of MX precipitates, while grain size is additional to the effect of MX precipitation in solution and is of secondary magnitude. In alloyed steels, especially Esshete 1250 austenitic steel, the effect of grain size is more difficult to obtain due to some precipitation occurring during creep exposure. Attention is given to the grain size evolution, and in particular, the size and distribution of MX (Nb-rich) particles. The size, distribution of MX precipitates was analysed with electron microscopy techniques, while standard metallographic grain evolution measurements in creep exposed samples was carried out. It is concluded that MX precipitation play an important role for the creep strength of Esshete 1250 under service conditions. This experimental work helped to establish knowledge of particle coarsening, together with modelling of the microstructural stability for the creep resistant alloy, in order to understand creep deformation metallurgy and to achieve more reliable creep resistant alloys.
10:45 AM - T1.3
High Temperature Corrosion of Alloy Inconel 617 in (He-CO-CO2) Environment with Varying Oxidation and Carburization Potentials.
Deepak Kumar 1 , Gary Was 1
1 Materials Science and Engineering, University of Michigan, Ann Arbor, Michigan, United States
Show AbstractThe reactive impurities, H2, H2O, CO, CO2 and CH4 in the coolant helium in very high temperature gas reactor (VHTR) can cause oxidation, carburization or decarburization of the heat exchanger metallic component. In order to determine the role of impurities, CO and CO2, simplified helium gas chemistries containing only CO and CO2 (ppm levels) in helium were selected for corrosion tests on alloy Inconel 617 at 850, 900, 950 and 1000 degree C. Partial pressure ratios of (CO)/(CO2) over the range 7.5 to 1320 were chosen to determine the corrosion behavior as a function of oxidizing and carburizing potentials. Corrosion coupons were exposed to the various gas chemistries for up to 1000 h. A discharge ionization detector gas chromatograph was used to continuously analyze the impurity levels both before and after the furnace to understand the various gas/metal interactions occurring at the surface. The microstructure stability and the surface scale on the exposed samples were analyzed using SEM/optical microscopy. The composition and structure of the surface scale were analyzed using EDS and XRD. Corrosion kinetics, measured by the weight change, and activation energy are coupled with the analysis of structure and composition of the oxide to understand the corrosion mechanism. The results demonstrate that corrosion of alloy 617 is dependent upon the (CO/CO2) ratios in impure helium. The rate of corrosion was found to increase with (CO/CO2) ratio. At all tested temperature, CO2 acts as an oxidizing species causing oxidation of the alloy, whereas the role of CO is found to be dependent upon its partial pressure and temperature. Above a critical concentration (PCO*), CO can cause simultaneous oxidation and carburization of the alloys, whereas rapid decarburization of the alloy occurs below this critical CO concentration.
11:00 AM - T1: Stuct1
BREAK
T2: Carbides
Session Chairs
Monday PM, November 26, 2007
Gardner (Sheraton)
11:30 AM - T2.1
Microstuctural Characterization of the Radiation Effects in ZrC, a Potential Material for Next Generation Nuclear Plants.
Gianguido Baldinozzi 1 2 , Dominique Gosset 2 1 , Mickael Dolle 2 1 , David Simeone 2 1 , Suzy Surble 1 2 , Lionel Thome 3
1 Matériaux Fonctionnels pour l'Energie, CNRS-CEA-ECP, Châtenay-Malabry France, 2 Matériaux Fonctionnels pour l'Energie, CNRS-CEA-ECP, Gif-sur-Yvette France, 3 CSNSM, CNRS-IN2P3, Orsay France
Show AbstractThe development of a new generation of nuclear reactors (Gen-IV), with improved thermodynamic yield and a reduction of waste production, makes necessary to consider materials able to withstand high operating temperatures. Transition metal carbides, like ZrC, are then under consideration. Despite their good thermal and neutronic properties, they have unfortunately a brittle mechanical behaviour.This is the reason why it is important to investigate the properties of these systems as a function of the grain size. Therefore, samples having micrometric and nanometric grain sizes were irradiated by low energy ions at room temperature to simulate their behaviour in a neutron flux. The irradiation effects in these materials were studied by grazing X-ray diffraction an transmission electron microscopy.
11:45 AM - T2.2
SiC, TiC and ZrC Nanostructured Ceramics: Elaboration and Potentialities for Nuclear Applications.
Yann Leconte 1 , Marc Leparoux 5 , Auregane Audren 1 , Isabelle Monnet 2 , Lionel Thome 3 , Anna Swiderska-Sroda 4 , Stanislaw Gierlotka 4 , Sophie Le Gallet 6 , Xavier Portier 7 , Marc Levalois 7 , Nathalie Herlin-Boime 1 , Cecile Reynaud 1
1 DSM/DRECAM/SPAM/LFP, CEA-CNRS, Gif sur Yvette France, 5 Laboratory for Materials Technology, EMPA, Thun Switzerland, 2 DSM/DRECAM/CIRIL, CEA-CNRS, Caen France, 3 CSNSM, CNRS-IN2P3, Orsay France, 4 IHPP, Polish Academy of Sciences, Warsaw Poland, 6 Institut Carnot de Bourgogne, CNRS, Dijon France, 7 SIFCOM, CNRS, Caen France
Show AbstractCarbide ceramics as SiC, TiC or ZrC are potential candidates for high temperature applications such as fourth generation nuclear plants because of their refractory or low activation under neutron irradiation properties. Nevertheless, the typical drawbacks of hard ceramics (brittleness) could limit their use in these applications. In order to overcome these problems, one possibility is to decrease the grain size down to the nanometric scale. Enhancement of the mechanical properties is actually expected in such nanostructured ceramics (ductility) and moreover, these nanomaterials could also take advantage of their strong grain boundaries density to withstand severe irradiation conditions. If one wants to quantify the expected enhancement of the properties, the first challenge that has to be faced is the elaboration of the nanostructured ceramics samples. That means being able to synthesize the pre-ceramics nanopowders in weighable amounts, and then finding an efficient way to sinter them aiming at the maximum densification together with avoiding grain growth.In this contribution, we present SiC, TiC and ZrC nanopowders synthesis by laser pyrolysis and inductively coupled plasma, together with their densification by different techniques (Hot Isostatic Pressing, Spark Plasma Sintering, High Pressure Flash Sintering). We also report the latest findings obtained on the behavior of SiC nanostructured ceramics under low energy ion irradiation.Raw micrometric SiC and ZrC powders were used as precursors in the inductively coupled plasma experiment. The production was as high as 1 kg.h-1, with nanograins ranging from 10 to 100 nm in size depending on the synthesis conditions. For the laser pyrolysis method, gaseous precursors (SiH4, C2H2) were used for SiC while liquid alkoxides precursors were used for TiC and ZrC respectively. For SiC, the production rate can reach 100 g.h-1 (laboratory scale) with grain sizes ranging from 10 to 50 nm with narrow size distribution. For TiC and ZrC nanopowders, the production rate is lower than for SiC because of the use of liquid precursors that leads to a worse yield. In this latter case, the carbide phase is obtained after carburization of the laser pyrolyzed TiO2 (or ZrO2) / free carbon nanocomposites. The final carbide nanograins size is in the 50 – 80 nm range. After sintering, the obtained pellets show different characteristics depending on the starting powder and the sintering technique. With the right sintering conditions, the densification reaches 95 % without any sintering additives, with no (or limited) grain growth and no modification of the crystalline structure. Concerning the properties of the obtained nanostructured ceramics, the SiC pellets, together with the as-synthesized nanopowders, were submitted to low energy ion irradiation in order to compare their behavior to conventional SiC materials.
12:00 PM - T2.3
Investigation using Density Functional Theory of the Behavior of Krypton and Xenon in Silicon Carbide.
Marjorie Bertolus 1
1 DEC/SESC/LLCC, Commissariat a l'Energie Atomique, Saint-Paul-lez-Durance France
Show AbstractDuring in-reactor irradiation actinide fission and alpha disintegration produce large quantities of rare gases, which have a significant influence on the structural and mechanical properties of nuclear fuels. It is therefore essential to get further insight into the behavior of rare gases in materials to understand the behavior of fuels and improve their performance. The incorporation sites and activation energies of rare gases determine their mobility in the material and the influence of temperature and defects on this mobility. It is then of major importance to evaluate these parameters.¶¶Ab initio investigations of materials are currently done mostly within the Density Functional Theory (DFT) framework. Current DFT approximations, however, do not enable a correct description of purely dispersive interactions, as was shown on rare gas dimers [1]. We have therefore recently evaluated the accuracy of DFT for the bonds formed between rare gases and open-shell atoms on small molecules containing rare gases, for which experimental data exist in the literature, and for which very precise post-Hartree-Fock calculations are feasible [2]. This investigation shows that standard DFT describes correctly the bonds formed by xenon and krypton with open-shell atoms, while the bonds formed by helium atoms, which are very weak, are much more problematic.¶¶We present here an investigation using the generalized gradient approximation of DFT of the incorporation of krypton and xenon in silicon carbide, a material considered as a constituent for nuclear fuel of Generation IV future reactors. We have calculated the energies of one Kr or of one Xe atom in various sites of a stoichiometric or near-stoichiometric SiC crystal. We have also determined the most probable migration pathways between these sites and the activation energies associated. The evaluation of these energies enables us to get further insight into the atomic scale mechanisms involved in the migration and diffusion of Kr and Xe in silicon carbide.¶¶[1] T. van Mourik, R. J. Gdanitz, J. Chem. Phys. 116, 9620 (2002)¶[2] M. Bertolus, V. Brenner, in preparation
12:15 PM - T2.4
Characterization of Point Defect Generation, Migration and Coalescence in Irradiated SiC by Atomistic Simulation.
David Farrell 1 , Noam Bernstein 2 , Wing Kam Liu 1
1 Department of Mechanical Engineering, Northwestern University, Evanston, Illinois, United States, 2 Center for Computational Materials Science, Naval Research Laboratory, Washington D.C., District of Columbia, United States
Show AbstractRenewed interest in nuclear power in the United States has prompted investigations into new reactor designs, resulting in a need to gain a greater understanding of the properties of the materials which are proposed for use in next generation nuclear reactors. This presentation will focus on preliminary results of large-scale empirical potential atomistic studies into the generation of point defect clusters in 3C SiC by particle irradiation and the evolution from point defect clusters to ‘voids’ on the atomic scale. Our working definition of ‘void’ will be explained in the context of small length-scale simulations. The determination of interstitial and vacancy diffusivities for the empirical potential employed and its impact on defect coalescence will be discussed. The characterization of initial damage states for given irradiation conditions will be presented and compared to previous work on ceramics and ceramic composites.
12:30 PM - **T2.5
On Modeling the Evolution of Radiation Damage in Silicon Carbide.
William Weber 1 , Fei Gao 1 , Ram Devanathan 1 , Yanwen Zhang 2 , Weilin Jiang 1 , In-Tae Bae 1 , Zhouwen Rong 1
1 Fundamental Science Directorate, Pacific Northwest National Laboratory, Richland, Washington, United States, 2 Environmental Molecular Sciences Laboratory, Pacific Northwest National Laboratory, Richland, Washington, United States
Show AbstractExperimental charged-particle irradiations and multi-scale computer simulations have been used to investigate the primary damage state and evolution of damage in silicon carbide as functions of temperature and charged-particle mass and energy. Atomistic simulations of energetic C and Si collision cascades, similar to those created by reactor neutrons, indicate that single interstitials, vacancies, antisite defects, and small defect clusters are produced. The point defects are dominated by close Frenkel pairs, and atomistic simulations indicate that the activation energies for recombination of most close pairs range from 0.24 to 0.38 eV, which suggest significant reduction in defect survivability at room temperature. Atomistic simulations have also determined that the activation energies for long-range diffusion of C and Si interstitials are 0.7 and 1.5 eV, respectively. Using these activation energies and ab initio results as input parameters, a kinetic Monte Carlo (MC) simulation model has been developed to study isochronal annealing of defects in SiC between cascade events. The defects are produced by a 10 keV Si cascades in a molecular dynamics (MD) simulation cell, and these defects are then accurately transferred to defect lattice sites in the Monte Carlo model to investigate defect recovery. By transferring defects states back and forth between the MD and MC environments, damage accumulation can be investigated as a function of temperature. Charged particle irradiations are often used to simulate radiation damage from neutrons and radioactive decay; however, at extreme charged-particle fluxes used in irradiation studies to simulate radiation damage in nuclear materials, the ratio of ionization rate to displacement rate can have a significant impact on observed temperature-dependent processes, which can affect both interpretation and model development. At high charged-particle fluxes, the defect recovery rates in SiC increase nearly linearly with the ratio of ionization rate to displacement rate. A fundamental understanding of these ionization effects is needed if charged particle irradiation results are to be used to develop predictive models of damage evolution in nuclear materials, such as SiC, as functions of time, temperature and dose rates.
T3: Inert Matrix Fuels and Wasteforms I
Session Chairs
Monday PM, November 26, 2007
Gardner (Sheraton)
2:30 PM - T3.1
Simulations of Radiation Damage Production and Evolution in Spinels.
Blas Uberuaga 1 , Dnyansingh Bacorisen 3 , Jonathan Ball 2 , Roger Smith 3 , Robin Grimes 2 , Kurt Sickafus 1
1 , Los Alamos National Laboratory, Los Alamos, New Mexico, United States, 3 , Loughborough University, Loughborough United Kingdom, 2 , Imperial College, London United Kingdom
Show AbstractOxide spinels are candidate materials for a number of nuclear applications, including as inert matrices for nuclear fuels and radio-frequency windows for fusion reactors. Thus, there is an impetus for improving the understanding of the response of spinels in a radiation damage environment. The growing power and reliability of atomistic simulations means that various aspects of radiation response on the atomic level can now be probed. In this work, we apply these simulation methods to the problem of defect production and kinetics in spinels.In this talk, we will describe atomistic simulations of damage production in a series of three spinels that exhibit differing tendencies for cation disorder or inversion: MgAl2O4, which is a normal (cation inversion i~0) spinel, MgGa2O4 (i~0.5), and MgIn2O4 (i~1). Using molecular dynamics, the response of each of these spinels was analyzed and compared for primary knock-on energies of up to 10 keV. We then used accelerated molecular dynamics methods to probe the long-time behavior of the point defects that govern the evolution of the damage produced in the collision cascades. Further kinetic Monte Carlo simulations were performed in order to understand the effects of inversion on the diffusion of point defects in MgGa2O4 spinel.Our results show that defect production in all three spinels is primarily characterized by split interstitial/crowdions and cation antisite or disorder defects. In the normal spinel, subcascade branching occurs, resulting in relatively isolated pint defects. In contrast, the inverse spinels exhibit denser cascades with a core damage region that, at times, consists of a partial transformation of the spinel structure towards a rocksalt structure. As for defect mobility, we examined the migration mechanisms of point defects. We compare and contrast these defects in the three spinels, both in their ideal i=0 state as well as in disordered states representative of natural inversion levels. We find that, overall, interstitials diffuse much faster than vacancies, except in the case of Mg point defects, for which the migration energies of the vacancy and interstitial are similar.The introduction of disorder greatly modifies defect mobility. For example, the local arrangement of cations around different oxygen sites can be very different, leading to strong tendencies for point defects to prefer certain sites. We explored the consequences of this preference by examining in detail oxygen vacancy diffusion as a function of inversion in MgGa2O4. In this talk, we will discuss the implications these atomistic simulation results have for radiation tolerance in these materials. In particular, we will focus on the role inversion plays in both defect production and subsequent defect recombination and aggregation and how inversion may modify radiation tolerance.
2:45 PM - T3.2
Characterization and Dissolution Behavior of Uranium Containing Zirconia-Magnesia Inert Matrix Fuel.
Kiel Holliday 1 , Ken Czerwinski 1
1 Chemistry, Harry Reid Center, University of Nevada Las Vegas, Las Vegas, Nevada, United States
Show AbstractAn inert matrix of ZrO2-MgO containing UO2 and a burnable poison, Er2O3, is evaluated for its chemical dissolution behavior under a range of conditions. The anticipated enhanced chemical stability of UO2 in the inert matrix is a crucial waste form characteristic which has not been determined. The mixed oxide inert matrix has properties that are superior to the single component system. Zirconia alone is an insufficient inert matrix due to its low thermal conductivity producing high centerline temperatures. Pure MgO is an unacceptable inert matrix due to the high rate of hydrolysis in coolant water resulting in swelling, cracking, and eventual dissolution of the fuel in the event of a cladding failure. The synthesis of the ceramic is performed through a precipitation under elevated pH. Pellets of this material are characterized using electron and x-ray interactions to identify phases present, phase composition, and microstructure using x-ray fluorescence, x-ray diffraction, microprobe, and XAFS. The ceramics contain a cubic zirconia phase that includes UO2, Er2O3, and a small amount of MgO. A second pure MgO phase is also present. The zirconia phase is expected to enhance the durability of the ceramics as a waste form. A soxhlet apparatus is used to determine the corrosion resistance of the material should a direct disposal scheme be adopted. Pellets are exposed to hot water for over two thousand hours and the mass and volume changes are measured. The kinetics of the MgO hydrolysis in the presence of varied amount of zirconia is evaluated. The durability and dissolution behavior of the materials are further examined at elevated temperatures and high acidity to explore a means of dissolving the material for recycling. A pressure vessel is also used to determine the dissolution rates in simulated reactor conditions, using high temperature high pressure water. The dissolution kinetics of the ceramic elemental components is evaluated for the differing conditions. The solubility of the fuel in reactor, reprocessing, and repository conditions is investigated in a manner to provide thermodynamic and kinetic data necessary for modeling.
3:00 PM - **T3.3
Radiation Induced Structural Changes in Normal Spinels.
David Simeone 1 2 , Gianguido Baldinozzi 1 2 , Dominique Gosset 1 2 , Leo Mazerolles 4 , Suzy Surble 1 2 , Lionel Thome 3
1 Matériaux Fonctionnels pour l'Energie, CNRS-CEA-ECP, Châtenay-Malabry France, 2 Matériaux Fonctionnels pour l'Energie, CNRS-CEA-ECP, Gif-sur-Yvette France, 4 Institut des Sciences Chimiques Seine-Amon, CNRS, Thiais France, 3 CSNSM, CNRS-IN2P3, Orsay France
Show AbstractIon-irradiation induces structural changes in many intermetallics and ceramics, but compounds belonging to the spinel structural type (AB2O4) are generally considered fairly resistant to this perturbation. Therefore, spinel compounds are of interest as incineration matrix for nuclear waste. Ion irradiation induces order disorder transitions in metallic alloys and in some ceramics. Under the effect of ion irradiation, different configurations can be observed. Spinels are a prototype system where these mechanisms can be tested. In normal spinels, the atoms in the cubic cell are in an approximately close-packed arrangement, where A and B are respectively tetrahedrally coordinated and octahedrally coordinated cations. Anions locate at the corners of edge-sharing octahedra and tetrahedra. Heating the sample, A and B cations interchange. Cations are ordered onto preferred sites at low temperatures and become increasingly disordered over the available sites at high temperatures. It is generally accepted that the normal pattern for disordering the spinel structure is to act on the A and B site populations going from normal to inverse structure (the A[AB]O4-type spinel).We will discuss experimental evidence obtained by Rietveld refinement of diffraction patterns collected on irradiated samples and by electron microscopy. A microscopic model can be derived, allowing the possibility to identify the pertinent order parameter and to predict the way these spinel structures behave.
3:30 PM - T3.4
Investigation on Hydrothermal Corrosion of MgO-Pyrochlore Composite as an Inert Matrix for Use in Light Water Reactors (LWRs).
Peng Xu 1 , Samantha Yates 1 , Juan Nino 1
1 Materials Science and Engineering, University of Florida, Gainesville, Florida, United States
Show AbstractA stable and reliable Inert Matrix (IM) with similar thermophysical and neutronic properties to UO2 is important for burning plutonium fuel while generating less radioactive waste in the nuclear reactors. Among many potential candidates for IM, MgO-based oxide fuels have attracted significant attention due to their favorable physical properties, including good thermal conductivity (7 W/m.K at 1273 K), high melting point (> 3000 K), and good radiation resistance. However, MgO has very poor hydration resistance which rules it out for use in Light Water Reactors (LWRs). Recently, an MgO-pyrochlore cercer composite has been proposed as a potential IM with adequate thermal conductivity and improved hydration resistance. In this study, Nd2Zr2O7 was selected as the pyrochlore phase and was synthesized by both solid state processing and sol-gel processing. MgO was obtained after calcination of MgCO3. The two powders were mixed by three methods including mortar and pestle grinding, magnetic bar stirring, and ball milling. Three different microstructures were achieved and their corrosion behaviors were investigated. Among the three mixing methods, ball milling produced a homogeneous mixing of Nd2Zr2O7 and MgO phases, and a grain size of ~ 1 μm for both phases after sintering. Hydrothermal corrosion tests were conducted in an autoclave filled with 200 ml of deionized water and heated at 300 °C and saturation pressure. A single phase MgO pellet was dissolved in less than an hour, while a pure Nd2Zr2O7 pellet showed excellent corrosion resistance with no mass loss for up to 10 days. The composite pellet with 60 vol% Nd2Zr2O7 lasted over 700 hours with approximate 20% mass loss. It was also found that the mass loss of the composite exhibits a linear relationship with the exposure time. The Arrhenius equation was used in calculating the apparent activation energy of the overall corrosion process for the composites with different amount of MgO including 60 vol%, 50 vol% and 40 vol%. The apparent activation energy increased from 40±2 kJ/mol to 52±2 kJ/mol as the amount of MgO decreased from 60 vol% to 40 vol%. These results point to the fact that the overall corrosion of the MgO-Nd2Zr2O7 composite is governed by multiple mechanisms, such as hydration reaction and grain boundary diffusion. We will finally show that the corrosion resistance can be further enhanced by enlarging MgO grain size (minimizing grain boundary) and uniformly dispersing Nd2Zr2O7 (maximizing hydration barrier).
3:45 PM - T3.5
Microstructure-Thermal Conductivity Relationships in MgO-Pyrochlore Cercer Composites for Inert Matrix Materials.
Samantha Yates 1 2 , Peng Xu 2 , Ken McClellan 1 , Juan Nino 2
1 Structure/Property Relations, Los Alamos National Laboratory, Los Alamos, New Mexico, United States, 2 Materials Science and Engineering, University of Florida, Gainesville, Florida, United States
Show AbstractInert matrix (IM) materials for nuclear fuel in light water reactors must meet several critical requirements that include high temperature stability, good irradiation behavior, high thermal conductivity, and hot water corrosion resistance. MgO is an attractive IM candidate due to its high thermal conductivity and its ability to be easily reprocessed; however, its hot water corrosion resistance is poor. A composite approach has been investigated to improve the hot water corrosion resistance of the MgO, while maintaining an adequate thermal conductivity. The components of this cercer composite include a pyrochlore phase that acts as a hydration barrier to prevent hot water corrosion and MgO to supply the path for thermal conduction. Based on neutronic simulations the ideal MgO—pyrochlore ratio was selected and various processing methods were investigated to produce a homogeneous dual-phase interpenetrating composite microstructure. The sintering kinetics of the MgO—pyrochlore composite has been investigated; and the effect of varying the sintering temperature, time, and rate on the composite microstructure will be discussed. The effect of the resulting microstructures on the thermal conductivity of the composite will then be presented.
T4: Fuels I
Session Chairs
Monday PM, November 26, 2007
Gardner (Sheraton)
4:30 PM - T4.1
Minor Actinide MOX Fuel Development for the GNEP Program.
Stewart Voit 1 , Kenneth McClellan 2 , Christopher Stanek 2 , Marius Stan 2
1 Actinide & Fuel Cycle Technologies, Los Alamos National Laboratory, Los Alamos, New Mexico, United States, 2 Materials Science Division, Los Alamos National Laboratory, Los Alamos, New Mexico, United States
Show AbstractOne of the thrust areas for the Global Nuclear Energy Partnership (GNEP) Program is the development of actinide bearing fuels for transmutation in a fast reactor. Mixed oxide (MOX) fuel has an extensive fast reactor irradiation history and thus is being considered for use as a baseline composition from which minor actinides can be incorporated for use as a transmutation fuel. However, there is very little minor actinide-bearing MOX (MA-MOX) irradiation performance data. Furthermore, there is limited MA-MOX fabrication experience in the US, thus an effort is underway to determine the viability of MA-MOX as a transmutation fuel through a series of fabrication and irradiation tests.The MA-MOX fuel development approach for the GNEP Program is divided into three areas: 1) surrogate MA-MOX development, 2) thermochemical modeling, and 3) MA-MOX development and fabrication. The surrogate development work uses actinide surrogates with depleted uranium to determine compositional dependent trends in the kinetics of sintering. In addition, a methodology is being established for determining the oxygen to metal (O/M) ratio of MA-MOX fuel. The modeling effort is focused on developing a practical thermochemical model of the O/M ratio as a function of temperature and oxygen partial pressure during sintering. This model will help to establish sintering furnace parameters needed for the fabrication of MA-MOX fuel with an O/M ratio that may vary between 1.90 and 1.98. The experience gained from the surrogate and modeling efforts will be used to refine process parameters for the fabrication of MA-MOX fuel for a series of irradiation tests. Details of the fabrication and irradiation test plan and results of the cooordinated MA-MOX development effort are discussed.
4:45 PM - T4.2
Properties and Cladding Interactions of Advanced Metal Alloy Transmutation Fuels with High Rare Earth Content.
John Kennedy 1 , James Cole 1 , Douglas Burkes 1 , Dawn Janney 1 , Andrew Maddison 1 , Cynthia Papesch 1
1 Nuclear Fuels and Materials, Idaho National Laboratory, Idaho Falls, Idaho, United States
Show AbstractThe product stream generated from the pyroprocessing of spent fuel during the fast reactor recycle stage of the Global Nuclear Energy Partnership (GNEP) closed fuel cycle may contain percentages of remnant rare earth fission products. As part of the continuing development of advanced transmutation fuels within the GNEP program, a series of actinide bearing metal alloys with lanthanide additions have been prepared for characterization before being subjected to irradiation testing. With the in-growth of fission products into the alloys during irradiation, rare earth compositions of up to 10 wt% could be produced at the high burnups considered for these fuels. Compared to earlier studies performed at this Laboratory, the U-Pu-Am-Np-Zr alloys fabricated for this campaign have a reduced Zr component of only 15-20 wt%, an increased U component to 40-60 wt%, and a lanthanide content of 1-2 wt% for the fresh fuel and up to 10 wt% representing the end-of-life fuel. We offer here results from studies on fresh cast and annealed samples intended to anticipate evolution of certain microstructure features in the irradiated fuel as well as the influence of high lanthanide fission product build-up on the thermo-physical properties of the fuels, particularly thermal conductivity. In addition, there is evidence that high lanthanide build-up in the fuel can enhance fuel-cladding-chemical-interaction (FCCI) leading to potential cladding failure. A series of diffusion couples have been constructed and annealed at temperatures between 500°C and 850°C to mirror steady state and transient state conditions. Scanning electron microscopy studies of the fuel cladding interfaces reveal a complex interaction zone.
5:00 PM - **T4.3
Influence of Phase Stability on Radiation Damage Properties: Plutonium-Gallium Alloys.
Steven Valone 1 , Michael Baskes 1 , Blas Uberuaga 1 , Richard Martin 2
1 Materials Science and Technology Division, LANL, Los Alamos, New Mexico, United States, 2 Theoretical Division, LANL, Los Alamos, New Mexico, United States
Show AbstractModeling cascade and fission damage evolution of actinide materials of all kinds is essential for understanding their aging characteristics. As an example of how exotic some of the damage evolution behavior can be plutonium-gallium (Pu-Ga) alloys are explored as an example. Aging emanates from the wide variety of spontaneous decay and fission products, which in the case of the Pu are such species as helium (He) and uranium, among others, as well as interstitials and vacancies. To aid in our understanding, the modified embedded atom method (MEAM) formalism [1, 2] has been applied to the Pu-Ga-He system [3–5]. The behavior of defects in the fcc (δ) phase of Pu-based materials is strongly influenced by the metastability of this phase. The influence of this metastability on minimum displacement threshold energy, point defect characteristics, defect transport, and He bubbles is delineated.Further the MEAM model for Pu metal is revised so that it more accurately captures the behavior of the Ziegler-Biersack-Littmark (ZBL) model of ion-ion interactions [6]. Two revision are tested with somewhat different stiffnesses in the 2-1000 eV range. The revised models show higher damage levels at 20 KeV than an earlier model, suggesting that the behavior of the models above 100 eV is dominating damage production, at least in the earlier stages of the cascade. All the models substantially agree in the low energy range whose properties are determined by the bulk properties of the fcc phase.Finally advanced concepts for construct MEAM potentials that can track phase changes [7] and approaches to describing mixed valence states of metal ions of Pu and U [8] are discussed.This work was performed at Los Alamos National Laboratory under the auspices of the U. S. Department of Energy, under contract No. DE-AC52-06NA25396.[1] M. I. Baskes, Phys. Rev. B 46, 2727 (1992).[2] M. I. Baskes, Mater. Sci. Eng. A261, 165 (1999).[3] M. I. Baskes, K. Muralidharan, M. Stan, S. M. Valone, and F. J. Cherne, J. Metals 55, 41 (2003).[4] S. M. Valone, M. I. Baskes, M. Stan, T. E. Mitchell, A. C. Lawson, and K. E. Sickafus, J. Nucl. Mater. 324, 41 (2004).[5] S. M. Valone, M. I. Baskes, and R. L. Martin, Phys. Rev. B 73, 214209 (2006) and references therein.[6] S. M. Valone and M. I. Baskes, J. Comp. Aid. Mater. Des., accepted (2007).[7] M. I. Baskes, S. G. Srinivasan, S. M. Valone, and R. G. Hoagland, Phys. Rev. B 75, 94113 (2007).[8] S. M. Valone and S. R. Atlas, Phil. Mag. 86, 2683 (2006).
5:45 PM - T4.5
Molecular Dynamics Study of Diffusional Creep in Nanocrystalline UO2.
Tapan Desai 1 , Paul Millett 1 , Dieter Wolf 1
1 Material Sciences, Idaho National Laboratory, Idaho Falls, Idaho, United States
Show AbstractWe have performed Molecular Dynamics (MD) simulations to study creep in nanocrystalline UO2 at temperatures well above the oxygen sub-lattice melting. In the absence of any external loading, we found that the uranium ions diffuse only via the grain boundaries (GB) whereas the oxygen ions show lattice as well as GB diffusion. When these microstructures are subjected to constant-stress loading at levels low enough to avoid microcracking and dislocation nucleation from the GBs, our simulations reveal that in the absence of grain growth UO2 deforms via GB diffusion creep (also known as Coble creep). The creep activation energy agrees well with the zero-stress diffusional activation energy of the slowest moving species, i.e. the uranium ions. Thus the rate-limiting mechanism for the Coble creep is the GB diffusion of uranium ions.
Symposium Organizers
Ram Devanathan Pacific Northwest National Laboratory
Robin W. Grimes Imperial College London
Kazuhiro Yasuda Kyushu University
Blas Pedro Uberuaga Los Alamos National Laboratory
Constantin Meis CEA, INSTN-UESMS
T5/E5: Joint Session: Modeling Defects in Nuclear Materials
Session Chairs
Tuesday AM, November 27, 2007
Constitution B (Sheraton)
9:30 AM - **T5.1/E5.1
Defects and Impurities in Nuclear Materials from First Principles.
Chu Chun Fu 1 , F. Willaime 1
1 DEN/DMN, SRMP-CEA, Sacaly, Gif sur Yvette France
Show AbstractFerritic steels play an central role in metallurgical and nuclear technology, in particularas structural materials for fission and future fusion nuclear reactors.Although their mechanical properties have been extensively investigated, little is knownabout the structural, electronic and magnetic properties at the origin of their macroscopicbehavior. First principles calculations within the Density Functional Theory (DFT) provide suchinformation at atomic scale, which is not directly accessible through experiments.However, the application of first principles studies in this field is rather new compared withother Solid State Physics and Material Science disciplines.The objective of this talk is to report key contributions of recent DFT studies which allowto reconcile theory and experiments:The determination of the 3D migration of self-interstitial atoms (SIAs) -- elementary defectscreated by irradiation -- induced an overall revision of the widely accepted picture of damageaccumulation under irradiation predicted by empirical potentials [1]. The coupled ab initio andmesoscopic kinetic Monte Carlo simulation provided strong evidence to clarify long-standingcontroversial interpretations of electrical resistivity recovery experiments concerning themobility of vacancies, SIAs, and their clusters [2]. The behavior of Carbon, one of theessential alloying element in steels, and of impurities such as Phosphorus which are responsibleof reactor pressure vessel (RPV) steels embrittlement will also be discussed in detail [3].[1] C. C. Fu, F. Willaime and P. Ordejon, Phys. Rev. Lett. 92, 175503 (2004)[2] C. C. Fu et al. Nature Mater. 4, 68 (2005)[3] E. Meslin et al. Phys. Rev. B 75, 094303 (2007)
10:00 AM - **T5.2/E5.2
Computer Simulation of Defect Properties in Irradiated Metals.
David Bacon 1 , Alexander Barashev 1 , Andrew Calder 1 , Yuri Osetsky 2
1 Engineering, University of Liverpool, Liverpool United Kingdom, 2 , Oak Ridge National Laboratory, Oak Ridge, Tennessee, United States
Show Abstract10:30 AM - T5.3/E5.3
Ab initio-based Radiation-induced Segregation Modeling in Fe-Ni-Cr Alloys.
Julie Tucker 1 , Todd Allen 1 , Dane Morgan 2
1 Nuclear Engineering & Engineering Physics, University of Wisconsin - Madison, Madison, Wisconsin, United States, 2 Materials Science and Engineering, Univerisity of Wisconsin - Madison, Madison, Wisconsin, United States
Show AbstractHigh concentrations of point defects, such as those created in radiation environments, can cause severe material degradation as they migrate and cluster. Radiation induced segregation (RIS), the process by which the local composition of an alloy is altered near point defect sinks, is a phenomenon that has concerned the nuclear industry for decades. While substantial progress has been made in the area of RIS prediction by empirical fitting, many questions remain about the diffusion mechanisms of point defects and how they are affected by local environment changes in multi-component alloys. This research uses ab initio methods to determine diffusion coefficients associated with both vacancy and interstitial migration in Ni rich fcc Fe-Ni-Cr alloys. We find that the alloy kinetics differs significantly from that predicted by simple extrapolation of empirical fits to high-temperature data. The calculated diffusion coefficients are used to parameterize a rate theory type RIS model.
10:45 AM - T5.4/E5.4
Microstructural Evolution Under High Flux Irradiation of Dilute Fe-CuNiMnSi Alloys Studied by Atomic Kinetic Monte Carlo Model – Effect of the Self Interstitials.
Christophe Domain 1 2 , Edwige Vincent 1 2 , Raoul Ngayam Happy 2 1 , Charlotte Becquart 2
1 MMC, EDF R&D, Moret sur Loing France, 2 Laboratoire de Métallurgie Physique et Génie des Matériaux, UMR 8517, Villeneuve d'Ascq France
Show AbstractT6/E6: Joint Session: Modeling Defects in Nuclear Materials II
Session Chairs
Tuesday PM, November 27, 2007
Constitution B (Sheraton)
11:30 AM - **T6.1/E6.1
Multi-Scale Simulations of Ion-Solid Interactions in SiC and GaN.
Fei Gao 1 , William Weber 1 , Haiyan Xiao 2 3 , Zhouwen Rong 1 , Yanwen Zhang 1 , Ram Devanathan 1 , Lumin Wang 2 , Xiaotao Zu 3
1 MS K8-93, Pacific Northwest National Laboratory, Richland, Washington, United States, 2 Department of Nuclear Engineering and Radiological Sciences, University of Michigan, Ann Arbor, Michigan, United States, 3 Department of Applied Physics, University of Electronic Science and Technology of China, Chengdu, Sichuan, China
Show AbstractRecent progress in multi-scale simulations of fundamental ion-solid interactions in SiC and GaN is discussed. Large-scale ab initio simulation methods (up to a few thousand atoms) have been developed for the study of ion-solid interactions in materials, and these methods are employed to investigate the properties of defects and defect clusters. Atomic structures, formation energies and binding energies of these small clusters are determined, and their relative stabilities are described. Furthermore, ab initio molecular dynamics methods have been used to calculate the threshold displacement energy surface and to simulate the primary damage states for the PKA (primary knock-on atom) energies up to 1 keV in SiC and GaN. These simulations provide significant insights into electronic effects on ion-solid interaction processes. Molecular dynamics (MD) have been used to simulate the high-energy ion-solid interactions with energies up to 50 keV, while kinetic Monte Carlo methods have been employed to investigate the recovery of defects during annealing, with input parameters determined by ab initito calculations and empirical potential MD simulations. A large number of 10 keV displacement cascades are randomly generated in a model crystal to simulate multiple ion-solid interactions and damage accumulation, as well as the mechanisms controlling the crystalline-to-amorphous transition. The amorphous-to-crystalline (a-c) transition has also been studied using MD methods, with simulation times of up to a few hundred ns, in 4H-SiC at temperatures between 1000 and 2000 K. Based on a model developed in the previous annealing simulations of 3C-SiC, the activation energy spectra for recrystallization along the three directions have been determined. In general, the activation spectra show that there are a number of activation energy peaks associated with different recrystallization processes. These activation energy values for full recrystallization are in the range of from 1.2 to 1.7 eV for the amorphous layers with the a-c interfaces along [-12-10] and [-1010] directions, and 1.1 to 2.3 eV for the amorphous layer with the a-c interfaces along [0001] direction.
12:00 PM - T6.2/E6.2
Defect Structure and Stability in Uranium and Zirconium Nitrides.
Robin Grimes 1 , Eugene Kotomin 2
1 materials, Imperial College London, London United Kingdom, 2 Joint Research Centre, Institute for Transuranic Elements, Karlsruhe Germany
Show AbstractActinide nitrides exhibit higher thermal conductivity and metal density than their corresponding oxides. Consequently they may be preferred as fuels in certain circumstances. Unfortunately, the basic defect behaviour of nitrides is not nearly as well understood as that of oxides. Since fuel performance depends on defect mediated processes this presents a problem in our being able to establish a predictive capability for nitride systems. Here we will consider the basic defect properties of uranium and zirconium nitrides as predicted by atomic scale computer simulations. In all cases, quantum mechanical simulation was employed, based on the density functional codes CASTEP and VASP. The atomic and electronic structures of basic vacancy and interstitial defects were studied and these were used to understand basic defect processes associated with disorder and nonstoichiometry. Predictions were successfully compared with experimental data were possible.
12:15 PM - T6.3/E6.3
Insights into Radiation Tolerance of Ceramics from Large Scale Molecular Dynamics Simulations.
Ram Devanathan 1 , William Weber 1
1 Fundamental Science Directorate, Pacific Northwest National Laboratory, Richland, Washington, United States
Show AbstractRadiation tolerant ceramics are needed to advance the utilization of nuclear energy to meet rising global energy demand. They have potential to meet the demands of radiation environments in applications as nuclear fuel and waste forms. Their discovery and development are hampered by a lack of fundamental understanding of the physics underlying radiation tolerance of ceramics. Several theories have been advanced in the literature based on structure, radius ratio of cations and ionicity or covalency of bonds. Most of these theories focus on defect production and accumulation processes only, but neglect in-cascade and thermal annihilation of defects. We have performed large scale molecular dynamics simulations of 30 keV U and Zr recoils in zircon (ZrSiO4), zirconia (ZrO2) and yttria-stabilized zirconia (YSZ) to understand the atomic-level mechanisms that contribute to radiation tolerance, particularly fast annihilation processes. Zircon is amorphized by irradiation, while zirconia and YSZ do not undergo radiation-induced amorphization. Our results reveal that dynamic defect annihilation is very effective at controlling defect accumulation in radiation tolerant materials. Based on our results, we will discuss a strategy for improving the radiation tolerance of ceramics.
12:30 PM - **T6.4/E6.4
Molecular Dynamics Simulation of Irradiation Induced Phase Transition in MgAl2O4.
Alain Chartier 1 , Tomokazu Yamamoto 2 1 , Kazuhiro Yasuda 2 , Constantin Meis 3 , Kenichi Shiiyama 2 , Syo Matsumura 2
1 , CEA-Saclay, Gif-Sur-Yvette France, 2 , Kyushu University, Fukuoka Japan, 3 , INSTN, Gif-Sur-Yvette France
Show AbstractSpinel MgAl2O4 is known to be highly radiation resistant: it may loose its crystalline structure (i.e. become amorphous) at cryogenic temperatures and at very high doses. The eventual amorphization is preceded by a phase transition towards a crystalline state, whose structure is analyzed whether to be of rock-salt type or a disordered spinel, depending on authors. Finally, the amorphous state may be re-crystallized when submitted to electron irradiations, both at the amorphous / crystalline interface by epitaxial growth and in the amorphous region by nucleation and growth.In the present study, the objective is to reveal this kinetics of phase transitions under radiation of MgAl2O4 using molecular dynamics (MD) simulation of Frenkel pair (FP) accumulation, starting whether from the normal (N) spinel or from the amorphous (A) spinel. Such a FP accumulation is designed to mimic the point defects accumulations resulting from displacements cascades in N spinel, or will be considered for mimicking the electron irradiation of the A spinel. Starting from N-spinel, the structure transits directly towards the rock-salt (NaCl) spinel for temperature lower than 600K. This transition is preceded by an intermediate disordered (D) spinel for higher temperature than 600K Amorphization has not been observed with increasing the dose up to 68 dpc (at 30K). The critical dose for NaCl-spinel transition increased as a function of temperature. It relies on spontaneous recombination mechanisms as the simulation time is too short for thermal diffusion to occur.Starting from the A-spinel, the FP accumulation induced a re-crystallization towards the NaCl spinel, with few defective atoms. The re-crystallization is not induced by any local heating since the target temperature for each simulation is strictly maintained and since the amorphous spinel is stable in temperature. The atomic displacements induced by electron irradiation can thus be considered as the driving mechanism that kinetically induces the re-crystallization. The super-cell used being small, such a re-crystallization can be consid-ered as homogeneous. As expected, the doses needed for the armorphous spinel to transit towards the rock-salt structure have been observed to decrease as a function of temperature. At 30K, a dose of around 6 dpc is needed for the re-crystallization to occur, while less than 1 dpc is needed at 1800 K.Finally, starting whether from N-spinel or A-spinel, FP accumulation drove the spinel towards the NaCl spinel structure (the steady state under irradiations), rather than the D-spinel. The transitions were driven by Frenkel-pairs recombination, as already observed in pyrochlore. We also demonstrated that non diffusive point defects (FPs, mimicking the electron irradiation) may induce temperature dependant homogeneous crystallization of A-spinel.
T7/E7: Joint Session: Modeling Microstructural Evolution in Irradiated Materials I
Session Chairs
Tuesday PM, November 27, 2007
Constitution B (Sheraton)
2:30 PM - **T7.1/E7.1
Atomistic Simulations and Continuum Modeling of Microstructural Evolutions Driven by Irradiation.
Pascal Bellon 1 , Pavel Krasnochtchekov 1 , Arnoldo Badillo 1 , Charles Enloe 1 , Robert Averback 1
1 Materials Science and Engineering, university of Illinois, Urbana, Illinois, United States
Show AbstractIrradiation drives materials into nonequilibrium states, resulting into forced phase transformations, microstructural evolutions, and dimensional changes. In advanced nuclear reactors, such as GEN IV fission reactors and fusion rectors, materials will be subjected to untested irradiation environments, in particular large doses and high irradiation temperatures. Due to time constraints, it is not possible to perform experiments in the actual service conditions these materials would be subjected to. Simulations and modeling will thus be employed to overcome this limitation, either by allowing a safe extrapolation of results obtained in accelerated tests, or even by brute force multi-scale modeling. As a first example, we will show how atomistic simulations (MD, KMC) and quantitative continuum modeling can be used to study the stability of pre-existing precipitates, or the formation of nonequilibrium precipitates during the irradiation of metallic alloys. We will show how the characteristic length scale of the atomic replacements forced by nuclear collision play a role on precipitate stability, and on the possible self-organization of these precipitates in nanoscale patterns. We will also show that, when interstitials dominate the transport of solute atoms, the resulting morphology of the precipitates is quite different from the case where solute transport is dominated by vacancies. In a second example, we will show how quantitative phase field modeling can be employed to simulate the evolution of point defect clusters under irradiation. For making safe predictions as discussed above, it is imperative that the primary state of damage be taken into account. This information, however, is absent from most existing models since it relates to the amplitude of the fluctuations of the external forcing. We have introduced a mixed discrete-continuum approach to overcome this problem, and we will show how the primary state of damage affects the size and number density of defect clusters. The results will be compared to existing simulation results obtained by MD+KMC and by cluster dynamics.
3:00 PM - T7.2/E7.2
Multi-Physics Simulation of Grain Restructuring in Fast Reactor Nuclear Fuels.
Veena Tikare 1 , Timothy Bartel 1 , Mark Lusk 2 , Steven Wright 1
1 , Sandia National Laboratories, Albuquerque, New Mexico, United States, 2 , Colorado School of Mines, Golden, Colorado, United States
Show AbstractFast reactor oxide nuclear fuels are cylindrical pellets that are approximately 1 inch in height by 1/4 inch in diameter. Their initial microstructure is homogeneous with equi-axed grains of 10 to 20 μm and uniformly distributed porosity of 10 to 15 vol%. During service in fast reactors, these fuel pellets experience high service temperatures with very large radial temperature gradients, which drive tremendous restructuring of the microstructure by grain growth and pore migration. Furthermore, volumetric strain, called swelling, by creep occurs to accommodate the addition atoms generated by fission. The restructuring, fission product release and migration modeling is required to accurately predict the performance behavior of fuel-cladding mechanical and chemical interactions that control the integrity of the fuel pin and its ability to operate as designed without releasing fission products. We will extend a current Material Point Method, MPM, model to incorporate kinetic Monte Carlo, kMC, algorithms to simulate fuel restructuring due to curvature driven grain growth and pore migration. We will also simulate swelling by creep. This multi-physics model will be presented, it application to simulating fuel restructuring will be demonstrated and finally the limitations of the model will be discussed.This work was done at Sandia National Laboratories, a multiprogram laboratory operated by Sandia Corporation, a Lockheed Martin Company, for the United States Department of Energy under Contract DE-AC04-94Al85000.
3:15 PM - T7.3/E7.3
Computer Simulation of Helium Gas Bubbles in Uranium Dioxide.
David Parfitt 1 , Robin Grimes 1 , Kaajal Desai 1
1 Department of Materials, Imperial College London, London United Kingdom
Show AbstractHelium is formed in nuclear fuel as a product of the α-decay of actinides during long-term storage and normal operation. The high heat of solution of this helium drives the precipitation of the atoms into gas bubbles; these act as reservoirs, absorbing helium from the lattice, and through thermal and radiation-enhanced resolution, returning it back to the lattice. Understanding the effect of these bubbles upon the thermal and mechanical properties of uranium dioxide is important for safe and economical storage and operation. Here we present molecular dynamics simulations of helium gas bubbles and their interaction with displacement cascades within the uranium dioxide lattice. A simple two-body effective potential has been used to successfully model the dynamics of several different bubbles sizes and morphologies with a range of internal gas pressures. The interaction of these bubbles with recoil cascades, representative of radiation damage due to alpha-decay and nuclear fission, has been examined with initial recoil energies of up to 30 keV. Our results show that, as expected, the thermal resolution of helium is almost non-existent at typical fuel storage and operating temperatures. In contrast to previous models however, the dominant mechanism of radiation-enhanced resolution is not the ballistic recoil of helium atoms, rather it results from the rapid and concerted incorporation of the helium into the disordered regions adjacent to the bubble. Once absorbed into the disordered region, this helium acts to prolong its lifetime far beyond that normally observed in such radiation damage simulations. We propose that these new mechanisms will significantly impact upon models of the gas resolution rate and the mobility of bubbles in uranium dioxide. In particular, the combination of re-absorption of helium into the disordered regions of the lattice and the transfer of uranium and oxygen atoms across the bubble will lead to the net migration of these bubbles along any anisotropy in the radiation field.
3:30 PM - T7.4/E7.4
Reactor Transient Material Instability Models for High Temperature and High Burn-up Nuclear Fuel.
Ray Stout 1
1 , Rho Beta Sigma Affaires, Livermore, California, United States
Show AbstractThe designs of next generation nuclear reactors that operate efficiently and perform safely at higher localized fuel temperatures and fuel burn-ups will require an increased mechanistic understanding of materials’ responses to develop and evaluate ceramic nuclear fuel performance. At higher temperatures, any given non-elastic/plastic deformation rate will occur at a lower stress state and the transport rate of fission gas atoms to bubbles will also be increased. The higher burn-ups, measured as fissions per unit volume of ceramic nuclear fuel, will locally increase the number density of fission gas atoms that will be transported to existing porosity and/or additionally created fuel bubbles. Based on these facts, pressurized bubble density in nuclear fuels and porosity swelling strains of future reactor designs, as measured by the number of bubbles per size attributes(radius and radius rate) and per fission gas(atom content), will evolve significantly different from the pressurized bubbles densities observed in the existing light-water reactors. The quasi-steady evolution of bubble density and bubble density dependent fuel deformation has been formulated for nuclear fuel response models[1] in terms of: (1). A bubble density field equation to describe the time evolution of the discrete bubble species of different size(radius) bubbles and different gas content bubbles; (2). finite deformation and finite strain dependence on fission gas bubble density; and(3). stress/bubble-pressure equilibrium dependence on fission gas bubble density. This system of equations is “quasi-rate unstable” because the material stiffness response decreases as the bubble density evolves. For reactor quasi-steady operation, this is a reactor life-time constraint. However, for any operational excursion from an existing fuel state (fuel state is characterized by fuel temperature and fuel burn-up), the safety of reactor performance is a decreasingly limited functional of fuel temperature and/or fuel burn-up excursion increments.[1]. Stout, R.B.[2006]: Stochastic Deformations and Bubble Density Evolution in Nuclear Materials, rbsA-Rpt0016, Jun06.
3:45 PM - T7.5/E7.5
Phase Field Modeling of the Effect of Irradiation Damage on Thermal Conductivity at the Microstructural Scale.
Paul Millett 1 , Michael Pernice 2 , Tapan Desai 1 , Dieter Wolf 1
1 Material Sciences Dept., Idaho National Laboratory, Idaho Falls, Idaho, United States, 2 Center for Advanced Modeling and Simulation, Idaho National Laboratory, Idaho Falls, Idaho, United States
Show AbstractT8/E8: Joint Session: Modeling Microstructural Evolution in Irradiated Materials II
Session Chairs
Tuesday PM, November 27, 2007
Constitution B (Sheraton)
4:15 PM - **T8.1/E8.1
Effect of Strain Field on the Microstructural Evolution in Irradiated Fe: Kinetic Monte Carlo Study.
Naoki Soneda 1 , Kenichi Nakashima 1 , Akiyoshi Nomoto 1 , Akiyuki Takahashi 2 , Toshiharu Ohnuma 1 , Kenji Nishida 1 , Kenji Dohi 1
1 Materials Science Research Laboratory, CRIEPI, Komae, Tokyo, Japan, 2 Department of Mechanical Engineering, Tokyo University of Science, Noda, Chiba, Japan
Show AbstractMicrostructural features in irradiated metals, such as dislocations, dislocation loops, voids and solute clusters / precipitates, cause strain fields in the crystal, which affect the diffusion of point defects and solute atoms. Understanding the effect of such strain field is very important to develop models for point defect cluster formation, dislocation decoration with SIA loops, solute segregation to grain boundaries and heterogeneous solute clustering. Kinetic (Lattice) Monte Carlo simulation is a powerful technique to describe the microstructural evolution such as point defect and solute clustering under irradiation. The conventional K(L)MC, however, does not consider the effect of strain field explicitly. Rather it employs the concept of capture radius for the interaction between particles, within which the interaction occurs spontaneously. Therefore it is likely that this conventional K(L)MC approach may not be appropriate to perform the simulations of especially the interaction of defects with dislocations and the effect of high doses where a lot of well-developed irradiation features exist.In this paper, we will present our recent results to consider the effect of strain field in the KMC simulations. In our approach, computation box is subdivided into small cells, and each cell is assigned an appropriate constant strain. Migration and formation energies of point defects and clusters are calculated as a function of strain, and the jump probability of a particle is calculated using the migration and formation energies corresponding to the strain value of the cell of the particle. Results of this new KMC simulation will be presented for the microstructural evolution in irradiated bcc Fe.
4:45 PM - **T8.2/E8.2
On the Role of Helical Turns in the Formation of Clear Bands in Irradiated Metals.
Thomas Nogaret 1 , Marc Fivel 1 , David Rodney 1
1 SIMAP, INP Grenoble, Saint Martin d Heres France
Show Abstract5:15 PM - T8.3/E8.3
Structure and Defect Stability of Calcium Phosphate Minerals.
Emily Michie 1 , Robin Grimes 1
1 Materials, Imperial College London, London United Kingdom
Show AbstractThe choice of material for radioactive waste immobilization depends both upon the chemistry of the waste type and the physical environment in which the waste form is to be stored. Natural mineral phases are attractive due to their proven long time scale stability. Consequently, apatite is of interest because (i) it is the most abundant phosphatic mineral, found in almost all igneous, some sedimentary and metamorphic rocks and (ii) it exhibits a large chemical variability.
Indeed, due to the difficulty of incorporating halides in conventional materials, apatites are being considered as potential hosts for radioactive waste streams containing both actinides and halides. Apatite can demonstrate considerable non-stoichiometry, providing significant compositional flexibility, ideal for incorporating a range of waste species. It is therefore imperative to better understand the stability of apatites and the relative solubility of different ions.
Atomistic scale computer simulation is used to examine the apatite structure and the defect mechanisms associated with incorporating both alkali halide and various cation species. Initially, the structures for various fluorapatites were established by comparing quantum based calculations (CASTEP) and experimental values.
Of particular interest was the substitution of Sr2+ onto the two non-equivalent Ca2+ cation sites in fluoroapatite. Previous experimental studies are unclear as to which site Sr2+ substitutions are most likely to occur. Here it is found that there is no energetic difference between substituting Sr2+ onto either site. Zn2+, Mg2+, and Ba2+ substitutions were also investigated, however these showed a significant difference between the cation sites.
5:30 PM - T8.4/E8.4
Dynamics of Irradiation-induced Nano-fiber Formation in Germanium.
Kun-Dar Li 1 , Wei Lu 3 , Lumin Wang 1 2
1 Department of Materials Science and Engineering, University of Michigan, Ann Arbor, Michigan, United States, 3 Department of Mechanical Engineering, University of Michigan, Ann Arbor, Michigan, United States, 2 Department of Nuclear Engineering & Radiological Science, University of Michigan, Ann Arbor, Michigan, United States
Show AbstractIt was initially thought that amorphous materials would be ideal for use in a radiation environment to avoid the additional disorder. However, irradiation induced catastrophic swelling has been observed in several amorphous materials, e.g. amorphous germanium, by experiments with ion beam irradiation in the past two decades [1-2]. A common ion irradiation effect in these materials is the development of a high density of porosity separated by nanofibers of ~10-20 nm in diameter. The mechanisms for the formation of this porous structure in the amorphous phase are still unclear. In this study, we develop a theoretical model that considers the kinetics of defect build-up during irradiation, including the recombination effect of defect and the dynamic diffusion process. A phase field approach is adopted in this model. Pores are treated as high vacancy concentration regions. The local change in defect concentration can be written as the net result of local production rate, reaction rates and divergence of diffusion flow. Dynamic processes, such as pores coalescence, are captured by updating the concentration profile over time. We incorporate the free energy of mixing and interfacial energy into the driving force for vacancy diffusion (based on Cahn-Hilliard equation) while the production and annihilation of vacancy that are due to the ion irradiation and the reactions with interstitials and sinks, respectively (based on Sizmann’s equation). The application of a diffuse interface allows pores to emerge or dissolve naturally, and the system can form whatever lattice it favors. A series of simulations are performed for dynamic porous formation and evolution by this model. It has been observed that initially spherical voids are formed and their size increases over time. Continuous observation of a set of voids in the simulation reveals void coalescence events during the early stage of pore growth. Eventually they become enlarged leading to a sponge-like structure. With increasing the irradiation dose, the structure continues to evolve into a network of fiber-like structure with nearly uniform diameters of fibers. These calculated evolutions of the porous structure are consistent well with all the experimental observations in different studies [2-4]. Our model provides a distinct numerical approach to investigate the mechanism of microstructural/morphological instability in irradiated germanium. The approach may also be applied to other irradiated materials such as GaSb and InSb.[1] B. R. Appleton, O. W. Holland, J. Narayan, O. E. Schow III, J. S. Williams, K. T. Short and E. M. Lawson, Appl. Phys. Lett. 41 (1982) 711.[2] L. M. Wang and R. C. Birtcher, Appl. Phys. Lett. 55 (24) (1989) 2494.[3] S. M. Kluth, J. D. Fitz Gerald and M. C. Ridgway, Appl. Phys. Lett. 86 (2005) 131920.[4] S. M. Kluth, David Llewellyn and M. C. Ridgway, Nuclear Instruments and Methods in Physics Research B 242 (2006) 640.
5:45 PM - T8.5/E8.5
Physics Mechanisms Involved in the Formation and Recrystallization of Amorphous Regions in Si through ion Irradiation.
Ivan Santos 1 , Luis Marques 1 , Lourdes Pelaz 1 , Pedro Lopez 1 , Maria Aboy 1
1 Departamento de Electronica, University of Valladolid, Valladolid, Valladolid, Spain
Show AbstractIon implantation and annealing are process traditionally used for the fabrication of Si devices. The introduction of energetic ions during the implantation step generates a large number of defects in the Si lattice. These defects may also diffuse, interact among them and with dopants, annihilate at interfaces, etc. resulting in a final dopant and defect configuration that determines the device performance. It is necessary have a good understanding of all these mechanism to optimize the device. Multi-scale modeling is required to capture the detailed physics involved in the mechanisms and to access to scales directly comparable with experiments. We focus this work on multi-scale modeling of the physics involved in ion beam induced amorphization and recrystallization in Si, but this scheme can be applied to other materials. We use molecular dynamics (MD) to study the formation mechanisms of amorphous regions. We have observed that along with energetic ballistic collisions that generate Frenkel pairs (energy transfers above the displacement energy in Si), low energy interactions can cause amorphous damage through the melting and quenching of local regions where energy is deposited. We will show that the competition between melting and heat diffusion define the conditions for the formation of local amorphous regions at the low energy regime (around and below the displacement energy). This mechanism of damage formation is very relevant for heavy and molecular ions.By quantifying the results obtained with MD, we have developed a model that complements the damage generation in the binary collision approximation (BCA). This model reproduces the characteristics of damage of MD cascades: amount and morphology of generated defects. Complex damage structures appear during the cascade generation in agreement with MD. However, this model has a much lower computational cost than MD simulations, which allows us to simulate thousand of cascades. We have successfully applied our model to B18 implantation.We have used MD results related to the recrystallization behavior of local amorphous regions to define the energetic of defects in a computationally efficient Kinetic Monte Carlo (KMC) code. The combination of all these tools, MD (fundamental studies of damage formation and recrystallization), improved BCA (including ballistic and melting-related damage) and KMC (for efficient defect kinetics modeling during the implantation and the subsequent annealing) allow us to model the effect of ion mass, beam current, implant temperature on the amount and morphology of residual defects in Si.
T9: Poster Session: Materials Innovation for Next-Generation Nuclear Energy
Session Chairs
Robin William Grimes
Blas Uberuaga
Wednesday AM, November 28, 2007
Exhibition Hall D (Hynes)
9:00 PM - T9.1
Oxidation Behavior of Zr-Nb Alloys at 973-1273 K in Air.
Tatsumi Arima 1 , Kensaku Miyata 1 , Kazuya Idemitsu 1 , Yaohiro Inagaki 1
1 , Kyushu university, Fukuoka Japan
Show AbstractIntroduction Oxidation behaviors of zirconium-niobium alloys have been extensively studied from aspects of thermal treatment, chemical composition, irradiation effect and so on. For the safe management of nuclear fuels, it is important to investigate oxidation behaviors under such LOCA conditions [1]. So far, we have reported oxidation behaviors of Zircaloy-2, -4 and Zr-Nb alloys under low oxygen potential below ca. 873 K and discussed the relationship among oxidation rate, structure of oxide film, oxygen partial pressure, etc. [2,3]. In the present study, oxidation behaviors of Zr-Nb alloys were investigated under the condition of high temperatures and dry air. Experimental In order to simulate an accident condition, oxidation tests were done at 973 K–1273 K under the flowing 21%O2-N2 gas. Temperature was achieved to annealing one within 120 s and was kept constant. For comparison, the flowing 1%CO-CO2 gas was used as well. Additive amount of Nb in Zr-Nb alloy varied from 1 to 10 wt%. In this study, Zircaloy-4 corresponds to 0 wt%Nb. Oxidation rates were measured by the thermo-balance. Before and after oxidation tests, all specimens were analyzed by the X-ray diffractometry (XRD). In addition, oxidized specimens were observed by an electron probe microanalyzer (EPMA).Results and discussion Under the 21%O2-N2 gas atmosphere, an amount of oxygen uptake increased with Nb content at 973 K and 1073 K, and the transition of oxidation rate from parabolic to linear was not observed. At 1173 K and 1273 K, such a rate transition was observed for large Nb content, and the amount of oxygen uptake was minimum at the vicinity of 2.5 wt%Nb. In addition, the amount of oxygen uptake for the 21%O2-N2 gas was much larger than that for the 1%CO-CO2 gas. XRD analyses showed that oxide film consisted of monoclinic ZrO2 and 6ZrO2-Nb2O5 for the condition of higher temperatures and large Nb content. On the other hand, at 973 K, tetragonal ZrO2 was observed together with m-ZrO2. The result obtained by EMPA showed that at high temperature of 1273 K, Nb precipitated at grain boundaries where existed widely around the oxide/metal interface.Summary Zr-Nb alloy showed oxidation resistance at 0-2.5 wt%Nb under the accident condition of high annealing temperatures and air. At 973 K and 1073 K, the oxidation rate for smaller Nb content was relatively small since tetragonal ZrO2 might play as a role of the oxygen diffusion barrier. Such an effect of oxidation retardation was reduced as temperature and Nb content increased. As a result, Zr-Nb alloys with large Nb content showed the large oxidation rate at higher temperatures.References[1] J.H. Baek, Y.H. Jeong, J. Nucl. Mater. 361 (2007) 30.[2] T. Arima, K. Miyata, Y. Inagaki, K. Idemitsu, Corr. Sci. 47 (2005) 435.[3] T. Arima, T. Masuzumi, H. Furuya, K. Idemitsu, Y. Inagaki, J. Nucl. Mater. 294 (2001) 148.
9:00 PM - T9.10
Progress in Out-Of-Pile Study of Low Enriched UMo/AlSi Fuel.
Marilyne Cornen 1 , Xaviere Iltis 1 , Fabrice Mazaudier 1
1 SPUA/LCU, CEA Cadarache, Saint Paul Lez Durance Cedex France
Show AbstractThe main problem we have to face in order to improve the in-pile behaviour of the UMo/Al dispersed fuel, is the instability of the interaction zone, that is formed between the particles and their surrounding matrix during irradiation. As the latest irradiations (e.g. IRIS 3 or RERTR 6) tend to prove, addition of Si into the Al matrix seems to be beneficial for the interaction rate decrease. To further understand the role of silicon, a wide out-of-pile interdiffusion study has been launched in CEA-France in 2006 and is still under progress. It has given promising preliminary results by showing a real effect of Si addition upon the interaction zone. These first results related to the physico-chemical aspects of the modified UMo/Al interaction as well as the kinetics aspects of the interaction layer (IL) formation are described in this paper. On a wide range of annealed couples made off UMo7 or UMo10 and Al-Si matrix (Si content ranging from 0.11 wt % to 12 wt %), the IL thicknesses have been systematically measured. Microstructures have been fully observed by means of optical and scanning electron microscopy and the elementary compositions have been evaluated through EDX. Finally, XRD characterizations and hardness tests have also been performed on a few samples. Complementary quantitative analyses through the IL by EPMA are scheduled by the end of 2007.On the basis of these different results, first clues about the role played by silicon on the UMo/Al interaction process are discussed.
9:00 PM - T9.11
Nanoindentation Studies of Ti, Zr, and Hf Hydrides.
Masato Ito 1 , Shunichirou Nishioka 1 , Hiroaki Muta 1 , Ken Kurosaki 1 , Masayoshi Uno 1 , Shinsuke Yamanaka 1
1 Division of sustainable energy and environmental engineering, Graduate school of engineering, Osaka University, Suita Japan
Show AbstractMetal hydrides, such as Ti, Zr, and Hf hydrides, are attractive and promising material as the U-Zr hydride fuel, neutron moderator, and reflector in nuclear plants. From the aspect of the integrity of structural materials, the several properties of hydrided metals have been extensively studied for a long time. However, the properties of the hydride itself have been scarcely studied due to the difficulty in production of bulk specimen, whereas the influence of the precipitated hydrides has been watched with keen notice. It is essential to study the thermophysical and thermochemical properties of single-phase metal hydrides in order to use them in practice. In this study, the nanoindentation studies on the Ti, Zr, and Hf hydrides were performed. Using a modified UHV Sieverts’ apparatus, the fluorite type structured single-phase bulk hydrides were successfully produced without any defects, and their mechanical properties such as elastic modulus and hardness were investigated. In addition, the microstructure and crystal orientation of the specimens were evaluated in order to discuss the elastic and plastic anisotropy of metal hydrides.
9:00 PM - T9.2
Study on Effects of Heavy Ion Irradiation on CeO2 by Using Synchrotron Radiation X-Ray Absorption Spectroscopy –As a Simulation Study for Radiation Damage in High-Burnup Light Water Reactor Fuels.
Akihiro Iwase 1 , Hirotaka Ohno 1 , Fuminobu Hori 1 , Norito Ishikawa 2 , Yuji Baba 3 , Norie Hirao 3 , Motoyasu Kinoshita 4 , Takeshi Sonoda 4 , Daiju Matsumura 3 , Yasuo NIshihata 3 , Jun'ichiro Mizuki 3
1 Department of Materials Science, Osaka Prefecture University, Sakai-shi Japan, 2 , Japan Atomic Energy Agency (Tokai), Tokai-mura Japan, 3 , Japan Atomic Energy Agency (Kansai), Sayo-cho Japan, 4 , Central Research Institute of Electric Power Industry, Komae-shi Japan
Show AbstractHigh burnup of light-water reactor (LWR) nuclear fuel (UO2) is one of good options in order to reduce the total amount of spent fuel and the fuel cycle costs. In high-burnup nuclear fuel pellets, however, a crystallographic re-structuring, which is called the rim structure is formed by high energy (around 100MeV) fission products (FPs). The rim structure may influence the fuel performance. Therefore, the mechanism of the rim structure formation has to be clarified in terms of fission product induced effects, i.e., high-density electronic excitation, defect production and inert gas accumulation in fuels. This study is performed to clarify the effects of FPs on nuclear fuels by using high energy ion irradiation and cerium dioxide (CeO2) with the fluorite structure as the simulation of high energy FPs and fluorite ceramics of UO2 fuel, respectively. To simulate the effects of defect production and inert gas accumulation, CeO2 pellets were irradiated with 3keV Ar ions in a vacuum chamber. Then, XPS spectra, and XANES spectra near Ar K-edge were measured in-situ. On the other hand, to simulate the electronic excitation effects of high energy FPs, CeO2 thin films were irradiated with 200MeV Xe ions at room temperature using the tandem accelerator at JAEA-Tokai. Structural changes of CeO2 by the irradiation were studied by using EXAFS spectroscopy around Ce K-edge at BL14 beamline of the synchrotron radiation facility, SPring-8 The behaviour of Xe atoms implanted near the surface of CeO2 was also studied by EXAFS around the Xe K-edge at SPring-8. XPS result shows that the charge state of Ce at the surface of CeO2 pellets changes from +4 to +3 by Ar irradiation. The reduction of Ce valence is accompanied by oxygen deficiency. XANES spectrum of Ar-irradiated CeO2 just above Ar K-edge does not show any structure for low irradiation fluence But for high fluence irradiation (1.0 x 1018 ions/cm2), the XANES spectrum shows an oscillation structure above the Ar edge. This result suggests that Ar atoms implanted in the matrix form solid aggregates under high pressure condition. In the symposium, results for EXAFS measurements around Ce and K-edges will also be discussed. The present experiment clearly indicates that the X-ray absorption spectroscopy by using synchrotron radiation facilities are quite useful for the study of irradiation effects of CeO2 , a simulation material of UO2 nuclear fuels pellets.
9:00 PM - T9.3
Mechanical Properties of Simulated ADS Target Fuel (UN+TiN (U:Ti=4:6) and U0.4Zr0.6N) and Inert Matrix (TiN and ZrN).
Jun Adachi 1 , Ken Kurosaki 1 , Masayoshi Uno 1 , Shinsuke Yamanaka 1 , Kazuo Minato 2
1 Division of sustainable energy and environmental engineearing, Osaka university, Osaka Japan, 2 Nuclear Science and Engineering Directorate, Japan Atomic Energy Agency, Osaka Japan
Show AbstractNitride inert matrix fuels are the candidates for ADS (Accelerator Driven System) target fuel. In the present study, we evaluated the mechanical properties of simulated ADS target fuels (UN+TiN(Ti/(U+Ti)=0.4) and U0.4Zr0.6N) and inert matrix (TiN and ZrN). The bulk samples of UN+60mol%TiN(92.8%T.D.), U0.4Zr0.6N(86.7%T.D.) and UN (93.8 %T.D.) were prepared by the carbothermic reduction. The bulk samples of TiN (94.0%T.D.) and ZrN (93.4%T.D.) were prepared by the spark plasma sintering. From XRD analysis and SEM-EDX, it was confirmed that ZrN formed the solid solution with UN and TiN precipitated in UN matrix. The mechanical properties of these samples were evaluated by Vickers hardness test, ultra sonic pulse echo method, and indentation technique. Young’s modulus of (U,Zr)N linearly increased with ZrN content and the hardness of U0.4Zr0.6N was higher than those of ZrN and UN. It is because the defect content of U0.4Zr0.6N is larger than those of UN and ZrN due to the large amount of the impurity oxygen. Young’s modulus and Vickers hardness of TiN phase in UN+TiN were lower than those of the single phase of TiN. On the other hand, those of UN phase in UN+TiN showed the similar values with the single phase of UN.Acknowledgment: This paper contains some results obtained within the task “Technological development of a nuclear fuel cycle based on nitride fuel and pyrochemical reprocessing” entrusted from the Ministry of Education, Culture, Sports, Science and Technology of Japan.
9:00 PM - T9.4
Study on Copper Precipitation under Electron Irradiation in FeCu Model Alloys by Using Vickers Hardness, Electrical Resistivity and Synchrotron Radiation X-Ray Absorption Spectroscopy.
Shou Nakagawa 1 , Fuminobu Hori 1 , Hirotaka Ohno 1 , Akihiro Iwase 1 , Ryuichiro Oshima 2 , Michiharu Kitagawa 2 , Norito Ishikawa 3 , Yoshihiro Okamoto 3
1 Department of Materials Science, Osaka Prefecture University, Sakai-shi Japan, 2 , Osaka Nuclear Science Association, Osaka, Osaka, Japan, 3 , Japan Atomic Energy Agency, Tokai-mura, Ibaraki, Japan
Show Abstract9:00 PM - T9.5
Thermodynamic Modeling of Plutonium Oxide Containing Americium.
Masayuki Hirota 1 , Ken Kurosaki 2 , Masayoshi Uno 2 , Shinsuke Yamanaka 2 , Shuhei Miwa 3 , Masahiko Osaka 3 , Kenya Tanaka 3
1 , College of Industrial Technology, Amagasaki, Hyogo, Japan, 2 , Graduate School of Engineering, Osaka University, Suita, Osaka, Japan, 3 , Japan Atomic Energy Agency, Oarai-machi, Ibaraki, Japan
Show Abstract9:00 PM - T9.6
Modeling and Simulation of Thermophysical Properties of Minor Actinides-Containing Oxide Fuels.
Masahito Katayama 1 , Jun Adachi 1 , Ken Kurosaki 1 , Masayoshi Uno 1 , Shuhei Miwa 2 , Masahiko Osaka 2 , Kenya Tanaka 2 , Shinsuke Yanamaka 1
1 Division of Sustainable Energy and Environmental Engineering, Graduate School of Engineering, Osaka University, Suita Japan, 2 , Japan Atomic Energy Agency, Higashiibaraki-gun Japan
Show AbstractThe molecular dynamics (MD) calculation was performed on minor actinides (MA)-containing mixed oxide (MOX) fuels, (U0.7-xPu0.3MAx)O2 (x = 0, 0.016, 0.03, 0.05, 0.1, 0.12, 0.15), and the lattice parameter, heat capacity and thermal conductivity were evaluated. The Morse-type potential function added to the Busing-Ida type potential was employed for the ionic interaction. The calculated lattice parameter increases with temperature and varied linearly with changes in the MA content. A marked increase in the heat capacity at constant volume is observed at around 2000~2500 K, which is probably caused by a presence of the Bredig transition. The calculated thermal conductivity data are comparable in (U0.7-xPu0.3MAx)O2 at any temperature and any x value, indicating that the thermal conductivity of MA-containing MOX fuels is scarcely influenced by adding MA up to 15 %. This result is reasonable because both the mass and size differences between uranium and neptunium and/or americium ions are very small. The present study shows that the MD calculation can be usefully applied to determine the thermophysical properties of MA-containing MOX fuels.
9:00 PM - T9.7
Evaluation of Mechanical Properties of Stabilized Zirconia and Zirconate Pyrochlore.
Takuya Matsuo 1 , Keiich Shimamura 1 , Tatsumi Arima 1 , Kazuya Idemitsu 1 , Yaohiro Inagaki 1
1 , Kyushu University, Fukuoka Japan
Show AbstractNuclear fuel cycle is expected to be safe, economical, ecological, sustainable, etc. Actinide doped zirconia-based oxides such as stabilized zirconia (Er0.05Y0.1PuzZr0.85-zO1.925) and zirconate pyrochlore (Zr2An2O7: An=actinide) are suitable materials to fulfill above requirements. So far, we have developed the manufacturing process of microspheres of stabilized zirconia added with cerium (surrogate for plutonium) [1]. Furthermore, its thermal properties have been systematically studied by molecular dynamics (MD) simulation [2]. In the present study, the mechanical properties were evaluated by experiment, MD simulation and ab initio calculation.In order to experimentally evaluate elastic moduli of zirconia-based oxides added with lanthanide (La, Ce, Nd, Sm, Gd) as surrogate for actinide (Pu, Am, Cm), ultrasound velocities in crystal were measured. MD calculations (MXDORTO program [3]) with Born-Mayer-Huggins interatomic potential were performed to obtain elastic moduli via the stress-strain relationship. Furthermore, ab initio calculations (WIEN2k code [4]) were performed using LAPW method with LSDA+U to obtain the bulk modulus via Birch-Murnaghan’s equation of state.Measurements of ultrasound velocities and MD calculations showed that bulk and Young moduli of stabilized zirconia added with Ce decreased with an increase of Ce content while the shear modulus was almost same. In addition, MD simulations showed that elastic moduli of Pu doped zirconia were a little larger than that of Ce doped one.Experimental result of zirconate pyrochlores showed that the elastic moduli increased with ionic radius of lanthanide. La has the largest ionic radius due to lanthanide contraction. As a result, La2Zr2O7 had the largest values of elastic moduli. Probably, actinide doped pyrochlore is anticipated to show the same tendency. MD simulations showed that the bulk modulus increased with the ionic radius whereas Young and shear moduli slightly decreased. As a result, the difference in elastic modulus between MD calculation and experiment became large as the ionic radius increased. Ab initio calculation showed that the bulk modulus of lanthanide pyrochlore roughly increased with ionic radius while that of actinide pyrochlore was almost same. MD simulation and ab initio calculation needed more careful treatment to precisely evaluate elastic moduli of pyrochlore.[1] T. Arima, K. Idemitsu, K. Yamahira, S. Torikai, Y. Inagaki, J. Alloys and Compd. 394 (2005) 271.[2] T. Arima, S. Yamasaki, K. Yamahira, K. Idemitsu, Y. Inagaki, C. Degueldre, J. Nucl. Mater. 352 (2006) 309.[3] K. Hirao, K. Kawamura, Material design using personal computer, Shokabo, Tokyo, 1994.[4] K. Schwarz, P. Blaha, G.K.H. Madsen, Comput. Phys. Commun. 147 (2002) 71.
9:00 PM - T9.8
Effects of Hydrogen Absorption on the Mechanical Properties of Zr-Nb Alloys.
Shunichiro Nishioka 1 , Masato Ito 1 , Hiroaki Muta 1 , Masayoshi Uno 1 , Shinsuke Yamanaka 1
1 Division of Sustainable Energy and Environmental Engineering,Graduate School of Engineering , Osaka University, Suita, Osaka, Japan
Show AbstractThe Zr-Nb alloys are expected for the new cladding tube materials and NDA, MDA, and ZIRLO have been newly used to improve corrosion resistivity and mechanical properties. However, burnup extension of the light-water reactors increases amounts of hydrogen absorption in the fuel cladding materials. The integrity of Zr-Nb alloys may be undermined when they absorb a lot of hydrogen. Therefore, it is important to investigate mechanical properties of hydrogenated Zr-Nb alloys. In this study, effects of hydrogen absorption on the mechanical properties of the cladding materials were evaluated. The hydrogenation of Zr-1.0Nb and Zr-2.5Nb was performed using a modified Sieverts’ UHV apparatus. The hydrogen contents of the specimens were measured by hydrogen analyser (HORIBA, EMGA-621).Young’s moduli of hydrogenated Zr-1.0Nb and Zr-2.5Nb were obtained by using a multiple elastometer (NTP, EG-HT) in the temperature range from room temperature to 773 K, based on the cantilever characteristic vibration technique. It was found that the solute hydrogen reduced the Young’s modulus of Zr-Nb alloys. The decreasing rate with hydrogen content was almost same as that of Zr hydrogen solid solution.
9:00 PM - T9.9
Thermophysical Properties of PuO2 and AmO2 Solid Solutions Simulated by Molecular Dynamics.
Tosawat Seetawan 1 , Thaweewat Khuangthip 1 , Vittaya Amornkitbamrung 2 , Ken Kurosaki 3 , Jun Adachi 3 , Masahito Katayama 3 , Anek Charoenphakdee 3 , Shinsuke Yamanaka 3
1 Physics, Sakon Nakhon Rajabhat University, Sakon Nakhon Thailand, 2 Physics, Khon Kaen University, Khon Kaen Thailand, 3 Division of Sustainable Energy and Environment Engineering, Osaka University, Suita Japan
Show AbstractPuO2 and AmO2 solid solutions: (Pu, Am)O2 are one of the candidates of fuels in a sub-critical accelerator-driven system (ADS). To understand the fuel performance, the thermophysical properties such as thermal conductivity and heat capacity are quite important. However, it is very hard to experimentally determine the physical properties of plutonium as well as americium compounds due to their handling-difficulties. Molecular dynamics (MD) would be a specific method to describe physical properties of such materials. In the present study, we have investigated thermophysical properties of PuO2, AmO2 and their solid solutions in the temperature range from 300 to 2500 K. The lattice parameter, compressibility, heat capacity, linear thermal expansion coefficient, and thermal conductivity were evaluated. A Morse-type potential function added to the Busing-Ida type potential was employed as the potential for interatomic interactions. The calculated lattice parameters of (Pu, Am)O2 obeyed Vegard’s law, and the values increased with temperature. The heat capacities of (Pu, Am)O2 were similar in any compositions. The thermal conductivities of (Pu, Am)O2 were lower than those of PuO2 and AmO2 , indicating that a point-defect scattering effect of phonons could be realized in the MD calculations.
Symposium Organizers
Ram Devanathan Pacific Northwest National Laboratory
Robin W. Grimes Imperial College London
Kazuhiro Yasuda Kyushu University
Blas Pedro Uberuaga Los Alamos National Laboratory
Constantin Meis CEA, INSTN-UESMS
T10: Structural Materials II
Session Chairs
Wednesday AM, November 28, 2007
Gardner (Sheraton)
9:45 AM - T10.1
Incoherent Interface Structure and the Fast Recombination of Irradiation-induced Frekel Pairs in CuNb Multilayer Composites.
Michael Demkowicz 1 , Richard Hoagland 1 , Amit Misra 2
1 MST-8, Los Alamos National Laboratory, Los Alamos, New Mexico, United States, 2 MPA-CINT, Los Alamos National Laboratory, Los Alamos, New Mexico, United States
Show AbstractAtomic-scale modeling using an EAM potential shows that incoherent FCC/BCC interfaces in CuNb multilayer composites exhibit a multitude of possible interfacial atomic configurations. This variety of arrangements has a profound effect on the behavior of interfacial point defects. In contrast to vacancies or interstitials in single crystals of Cu or Nb, point defects that have been absorbed at a CuNb interface do not remain compact and well-localized. Instead, they spread out over an area of about 30nm^2 within the interface by transforming the interface structure in the neighborhood of the site at which they were absorbed. Since the effective size of such interfacial defects is notably larger than that of the corresponding defects in single crystal Cu or Nb, the critical distance for Frenkel pair recombination is significantly increased at CuNb interfaces. Random walk simulations are used to demonstrate that this phenomenon predicts a reduction of radiation damage accumulation in CuNb multilayer composites, in agreement with experimental findings. This work was supported by the Los Alamos National Laboratory Directed Research and Development Program (LDRD) and the Los Alamos National Laboratory Director’s Fellowship Program.
10:00 AM - **T10.2
Light Water Reactor Fuel Degradation Mechanisms at High Burnup: Implications to Generation IV Materials.
Arthur Motta 1
1 , Penn State University, University Park, Pennsylvania, United States
Show AbstractThe proposed designs of GenIV reactors call for high operating temperatures and long exposure times, aimed at increasing thermal efficiency, enabling hydrogen production, decreasing waste and increasing safety, among other goals. These requirements will require imply the materials will be subjected to higher radiation damage doses than seen in light water reactors, requiring an even higher degree of alloys stability. In addition the higher temperatures and exposure times in corrosive environments will require much higher corrosion resistance than current materials. One of the main challenges of designing these materials is that our ability to predict material response in the face of the synergistic effects of temperature, radiation damage and a corrosive environment is limited. However, efforts are underway to understand mechanistically the degradation processes so that they can be extrapolated to conditions beyond the experimental database.In this context, it is interesting to consider the effort made by the nuclear power industry to qualify nuclear fuel for operation at higher burnup in existing light water reactors. In the last two decades, the average discharge burnup of nuclear fuel in light water reactors has increased almost two fold, thereby increasing the reactor exposure time and the amount of radiation damage withstood by the cladding. Experience has shown that the material degradation rates can increase at high burnup, sometimes through the synergistic operation of various processes. Two examples will be given to illustrate the complexity of the problem: (i)The rate of irradiation growth of zirconium alloys increases significantly at high burnup. The mechanism has been shown to depend on the amorphization of intermetallic ZrCrFe precipitates with consequent release of Fe into the Zr matrix. This has been shown to help nucleate component dislocation loops which preferentially absorb vacancies, thereby increasing the net rate of growth. This causes the interstitial and vacancy fluxes both to contribute towards irradiation growth, thus increasing the rate.(ii)One of the most significant obstacles to approval of cladding operation at high burnup is difficulty in the evaluation of the behavior of the fuel during a reactivity initiated accident (RIA). The degradation of the mechanical properties of the zirconium cladding with increasing corrosion and consequent hydrogen ingress, the radiation damage and the microstructure evolution in the fuel all have to be taken into account in evaluating the likelihood of severe fuel failure at high burnup. This problem is still being addressed in various research programs.Such mechanisms and the efforts to qualify nuclear fuel for higher burnup in light water reactors will be reviewed as a means of illustrating the challenges (known and unknown) faced by GenIV reactor materials.
10:30 AM - T10.3
In-situ X-ray Studies of Alloy Corrosion in Supercritical Water.
Kee-Chul Chang 1 , Bilge Yildiz 2 , Hamdallah Bearat 3 , Micheal McKelvy 3 , Dean Miller 1 , Hoydoo You 1
1 Materials Science Division, Argonne National Laboratory, Argonne, Illinois, United States, 2 Nuclear Engineering Division, Argonne National Laboratory, Argonne, Illinois, United States, 3 , Arizona State University, Tempe, Arizona, United States
Show AbstractThe main advantages of the supercritical-water cooled reactor (SCWR) are the simplicity of design due to the absence of coolant phase transitions and high thermal efficiency due to the high-operating temperatures. However, the high reactivity of the supercritical water can corrode even the most tenacious containment alloys such as stainless steel, which poses numerous materials science challenges under the extreme conditions of high pressure and high temperature. Therefore, a fundamental understanding of oxidation and corrosion at the interfaces of alloys and supercritical water, at a pressure and temperature higher than 218 atm and 374°C, is needed to develop alloys with superior durability for SCWR. We have conducted experiments to probe in-situ the initial stages of corrosion in Ni and stainless steel alloys, using a high-pressure cell with moissanite windows which allowed sufficient transmission of 25keV x-rays. We used x-ray reflectivity and diffraction to investigate the formation of new phases due to oxidation and corrosion at supercritical water conditions. We report our preliminary experimental results with this setup.
10:45 AM - T10.4
Thermal Helium Desorption of Helium-Implanted Iron.
Donghua Xu 1 , Kevin Wong 1 , Brian Wirth 1
1 Dept. of Nuclear Engineering, Univ. of California at Berkeley, Berkeley, California, United States
Show AbstractThe significant amount of helium produced in fusion environments can cause severe degradation of mechanical properties of the structural materials. Understanding the behavior of helium and its interactions with various micro/nano-structural features is one of the key issues in fusion reactor materials research. Thermal desorption experiments can provide important information about the kinetics and energetics of helium, which, when combined with multi-scale modeling and other specialized microstructural characterization, can disclose the atomistic mechanisms controlling He redistribution and bubble nucleation, important for developing a predictive model for the life performance of fusion reactors. Nevertheless, the execution of the experiments and the interpretation of the desorption data are not straightforward. In this work, we present our desorption data for single and polycrystalline bcc iron implanted with,4He or 3He at implantation energies from 5 to 100 keV and doses of 1013 - 1015 He/cm2. We also present an integrated numerical data analysis incorporating diffusion, detrapping, retrapping kinetics with useful input from other characterization experiments such as neutron depth profiling and positron annihilation spectroscopy that enable fitting of the data to deduce activation energies and characteristic pre-factors. Our results are compared with previous experimental and computational results.
11:00 AM - T10: Struct2
BREAK
T11: Nanomaterials
Session Chairs
Wednesday PM, November 28, 2007
Gardner (Sheraton)
11:30 AM - **T11.1
Radiation Damage Effects in Nanocrystalline Materials.
Kurt Sickafus 1
1 , los alamos national laboratory, Los Alamos, New Mexico, United States
Show Abstract12:00 PM - T11.2
Radiation Damage in Nanocrystalline UO2.
Dilpuneet Aidhy 1 , Tapan Desai 2 , Taku Watanabe 1 , James Tulenko 3 , Dieter Wolf 2 , Simon Phillpot 1
1 Department of Materials Science and Engineering, University of Florida, Gainesville, Florida, United States, 2 Material Sciences Department, Idaho National Laboratory, Idaho Falls, Idaho, United States, 3 Department of Nuclear and Radiological Engineering, University of Florida, Gainesville, Florida, United States
Show AbstractDisplacement cascades in nanocrystalline UO2 are studied using large-scale molecular dynamics (MD) simulation. A heavy cation (Uranium) is used as a primary knock-on atom (PKA) with energies ranging from 1 keV to 20 keV. Large number of point defect formation from radiation damage in single-crystal UO2 have been studied previously [1, 2]. To understand the ion irradiation damage at the grain boundaries, a nanocrystalline material consisting of three dimensional grains of 10-40 nm diameter size is used. Due to the radiation, we expect large-scale defect formation in the grains followed by segregation of these defects to the grain boundaries. We compare the effects of different interatomic potentials for the description of the interactions in UO2 on the defect evolution. The effect of the temperature on the defect formation/diffusion is also characterized. This work was supported by DOE NERI contracts DE-FC07-07ID14833 and DE-FC07-05ID14649.1.L.V. Brutzel, M. Rarivamanantsoa and D. Ghaleb, Journal of Nuclear Materials, 354, (2006), 28.2.T. Watanabe, S. Srivilliputhur, B. Uberauga, J. S. Tulenko, R. W. Grimes and S. R. Phillpot, (unpublished).
12:15 PM - T11.3
Atomistic Simulations of Diffusional Creep in Nanocrystalline BCC Molybdenum.
Paul Millett 1 , Vesselin Yamakov 2 , Tapan Desai 1 , Dieter Wolf 1
1 Material Sciences Dept., Idaho National Laboratory, Idaho Falls, Idaho, United States, 2 , National Institute of Aerospace, Hampton, Virginia, United States
Show AbstractThe high operating temperatures that exist in nuclear reactors denote that creep deformation is a critical process for both the nuclear fuel as well as the structural materials. In this work, molecular dynamics (MD) simulations are used to study diffusion-accomodated creep deformation in nanocrystalline molybdenum, a body-centered cubic metal commonly used as structural cladding. In our simulations, the microstructures are subjected to constant-stress loading at levels below the dislocation nucleation threshold and at high temperatures (i.e., T > 0.75Tmelt) thereby ensuring that the overall deformation is indeed attributable to atomic self-diffusion. In order to prevent grain growth and thus achieve steady-state creep, the initial microstructures were designed to consist of hexagonally-shaped columnar grains of uniform size and shape bounded by high-energy asymmetric tilt grain boundaries (GBs). The simulations are used to characterize the creep rates due to varying stress, temperature, and grain diameter. Remarkably, the results show that both GB diffusion in the form of Coble creep and lattice diffusion in the form of Nabarro-Herring creep contribute to the overall deformation. Visual analysis provides evidence that the GBs serve as sources for lattice vacancies that form within the GBs and subsequently emit into the grain interiors thus enabling lattice diffusion.
12:30 PM - T11.4
Radiation Damage Tolerance of Nanolaminate Composites.
Amit Misra 1 , Khalid Hattar 2 , Michael Demkowicz 2 , Richard Hoagland 2
1 Center for Integrated Nanotechnologies, Los Alamos National Laboratory, Los Alamos, New Mexico, United States, 2 Materials Science and Technology Division, Los Alamos National Laboratory, Los Alamos, New Mexico, United States
Show AbstractThis presentation will highlight experimental work on the development of radiation-damage-tolerant, ultra-high strength nanolaminate composites. The combination of high strength and high radiation damage tolerance obtains when the individual layers in these composites are only a few nanometers thick and therefore they contain an enormous interfacial area. These interfaces act both as obstacles to slip, as well as sinks for radiation-induced defects. Sputter-deposited Nb-Cu nanolaminates, with individual layer thicknesses varying from 4 nm to 100 nm, were used as model systems to explore the effect of the layer thickness on the radiation damage evolution due to 33 keV Helium ion implantation to doses ranging from 5 x 1016 /cm2 to 1.5 x 1017 /cm2. Implantation temperatures were varied from ambient to 500 °C. The samples were analyzed in their as-implanted state as well as after post-implantation annealing via Rutherford backscattering spectroscopy, elastic recoil detection and high-resolution transmission electron microscopy. Nanolaminate composites, presumably due to the large interfacial area per unit volume, exhibited enhanced solubility of He as compared to monolithic metals. Precipitation of nanometer-scale bubbles, preferentially along the interfaces in the 4 nm layers, was observed only at the highest implantation dose. During post-implantation annealing, nanolaminates with the 4 nm layers showed a suppression of bubble growth and He depletion, while retaining the layered morphology. For layer thickness of around 40 nm and higher, significant blistering and He depletion was observed. In the multilayers implanted at 500°C, the thicker layer thickness samples (100 nm) exhibited 20-50 nm size faceted bubbles in Cu and 1-2 nm size bubbles in Nb, while in the finer layer thickness samples (4 nm), the bubble size in Cu was limited by the layer thickness.The tailoring of layer dimensions and interface structures to design radiation-damage-tolerant nanolaminate composites will be discussed. This research is supported, in part, by the DOE, Office of Science, Office of Basic Energy Sciences.
12:45 PM - T11.5
Characterization of Nanodeposited Surfaces and Their Consequence on Critical Heat Flux Values.
Thomas McKrell 1 , S. Kim 1 , B. Truong 1 , L. Hu 2 , J. Buongiorno 1
1 Nuclear Science and Engineering, MIT, Cambridge, Massachusetts, United States, 2 Nuclear Reactor Laboratory, MIT, Cambridge, Massachusetts, United States
Show AbstractNanofluids, stable colloidal suspensions of nanoparticles, have shown great promise in improving the heat transfer characteristics of engineering systems such as nuclear power reactors. Specifically, it has been observed experimentally that nanofluids improve the maximum heat flux that a surface can transmit under boiling conditions, i.e. critical heat flux (CHF). It is shown that CHF enhancement results from the deposition of nanoparticles onto the heated surface not from unique thermal properties of the nanofluid. This finding required the use of techniques such as confocal microscopy, dynamic light scattering, scanning electron microscopy, energy dispersive X-ray spectroscopy, contact angle measurement, surface tension measurement, and X-ray diffraction. An emphasis is placed on quantification of key parameters such as surface morphology (surface roughness and area) and surface wettability and how these correlate to the observed CHF enhancement. The analytical techniques used to quantify the surfaces and aqueous nanoparticle suspensions tested, CHF experimental results obtained in our lab, and potential nuclear applications will also be discussed.
T12: Inert Matrix Fuels and Wasteforms II
Session Chairs
Wednesday PM, November 28, 2007
Gardner (Sheraton)
2:30 PM - **T12.1
Materials Innovation for the ``Back-End" of the Nuclear Fuel Cycle.
Rodney Ewing 1
1 Geological Sciences, University of Michigan, Ann Arbor, Michigan, United States
Show Abstract3:00 PM - T12.2
Characteristics of Radiation-induced Defects in Fluorite Structure Oxides Formed by Electron Irradiation.
Kazuhiro Yasuda 1 , Kazufumi Yasunaga 1 , Syo Matsumura 1 , Takeshi Sonoda 2
1 Appled Quantum Phys. and Nucl. Eng., Kyushu University, Fukuoka Japan, 2 Nuclear Technology Research Laboratory, CRIEPI, Komae Japan
Show AbstractOxide ceramics with fluorite structure attracts considerable attention as a host of inert matrix fuels and a transmutation target for minor actinides and long life fission products. We will report the characteristic features of defect clusters in oxide ceramics with fluorite structure, such as yttria stabilized zirconia (YSZ: 13 mol% Y2O3-ZrO2), ceria (CeO2), which were investigated in situ by transmission electron microscopy under electron irradiation. Results on calcium fluorite (CaF2) will be also shown for a comparison [1].Electron irradiation with 100 to 1000 keV was found to induce large defect clusters in YSZ, which accompany strong black/black lobes contrast and multiply dislocation networks when they grow to a critical diameter of about 1.0 μm. The defect clusters were considered to be charged oxygen platelets formed by the selective displacement damage of oxygen sublattice. A large difference in mass of constituent ions (O and Zr/Y ions) attributes the selective displacement damage under electron irradiation [2]. Similar defect clusters with strong diffraction contrast were formed in CaF2 under 200 keV electron irradiation, and they also multiplied dislocations at a critical diameter [1]. In the case of CeO2, a higher density of defect clusters was formed compared with YSZ and CaF2. Defect clusters formed under electron irradiation less than 1250 keV were found to be faulted-prismatic-interstitial loops lying on {111} planes with Burgers vector parallel to <111> directions [3], whereas electron irradiation with 1500, 2000 and 3000 keV induces perfect dislocation loops of 1/2<110>{110} in nature. Formation of radiation-induced defects is found to depend on electron energy for crystals with fluorite structure investigated in this study. Results will be discussed through selective displacement damage and charge accumulation.This work was financially supported by the Budget for Nuclear Research of MEXT, based on the screening and counseling by the Atomic Energy Commission, and by the Grant-in-Aid for Scientific Research (B) from JSPS. [1] M. Watanabe, T. Noma, K. Yasuda, K. Yasunaga, S. Matsumura and C. Kinoshita, Proc. of 16th Electron Mmicroscopy Congress, (2006) 1854.[2] K. Yasuda, C. Kinoshita, S. Matsumura, A.I. Ryazanov, J. Nucl. Mater., 319 (2003) 74.[3] K. Yasunaga, K. Yasuda, S. Matsumura and T. Sonoda, Nucl. Instr. and Meth. B, 250 (2006) 114.
3:15 PM - T12.3
Ion Irradiation Damage Effects in δ-phase WY6O12 and WYb6O12 at Cryogenic Temperature.
Ming Tang 1 , James Valdez 1 , Kurt Sickafus 1
1 , Los Alamos National Laboratory, Los Alamos, New Mexico, United States
Show Abstract3:30 PM - T12.4
Study of Fission and High-burnup Induced Restructuring of Nuclear Fuel Ceramics - Applying Computer Science to Investigate Kinetic Process.
Moto-yasu Kinoshita 1 2 4 , Y. Chen 2 , Y. Kaneta 2 , H. Geng 2 , M. Iwasawa 1 , T. Ohnuma 1 , T. Ichinomiya 3 , Y. Nishiura 3 , M. Itakura 4 , J. Nakamura 4 , K. Misoo 5 , S. Suzuki 5 , H. Matzke 6
1 , CRIEPI/JAEA/University of Tokyo, Tokyo Japan, 2 , University of Tokyo, Tokyo Japan, 4 , Japan Atomic Energy Agency, Tokai-mura Japan, 3 , Hokkaido University, Sapporo-shi Japan, 5 , Information and Mathematical Science Laboratory Inc., Tokyo Japan, 6 , Academy of Ceramics, Tokyo Japan
Show Abstract3:45 PM - T12.5
A Fundamental Study for the Understanding of Inorganic Iodine Waste Forms.
Tina Nenoff 1 , James Krumhansl 2 , Huizhen Gao 3 , Ashwath Rajan 2
1 Surface & Interface Sciences, Sandia National Labs, Albuquerque, New Mexico, United States, 2 Geochemistry, Sandia National Labs, Albuquerque, New Mexico, United States, 3 Radiological Consequence Management , Sandia National Labs, Albuquerque, New Mexico, United States
Show AbstractIodine waste forms are being developed from the leading ceramic components found in the DOE and open literature. Criteria of the waste forms include low cost fabrication, ease of fabrication, loading levels and durability of waste form. The waste form development is directed toward integration with the Yucca Mountain acceptance criteria. Specialized material science synthesis and state of the art characterization techniques are employed for rapid materials development and optimization. In particular, we are focused on two ceramic families of compounds that have high loading of iodine, plus have high mechanical, chemical and thermal stability with time. The first class of materials is metal-doped aluminosilicates, silicates and amorphous silicas; the second class is bismuth-based compounds. First, we focus on the development of optimized ceramic “Zeolite-Ag-I”-based waste form with a full understanding of its durability. This includes a full characterization and description of the ceramic(s) that result from various zeolites, iodine loadings and heat treatment processes. Second, we focus on the structure/property relationship between radioiodine and potential inorganic waste form storage phases. Novel Bismuth compounds being developed in our labs are potential low cost alternative waste forms. Two bismuth-iodide-oxides (“Phase 1”, Bi:I = ~1.9, “Phase II”,Bi:I ~4.6) were prepared from caustic iodine capture solutions. They are directly compared to our heat-treated zeolite-Ag-I waste form phases. Results are presented in detail; to date the bismuth compounds appear to offer a promising alternative for sequestering and storing radioiodine.A number of characterization techniques have been employed to better understand the relationship between the phases and their abilities to capture and retain Iodine, and their stability under possible repository conditions. They include, leach testing with elevated levels of common ground water anions (ie., choride and carbonate), thermal stability testing, elemental analysis, powder X-ray diffraction, SEM/TEM and FTIR studies. Sandia is a multiprogram laboratory operated by Sandia Corporation, a Lockheed Martin Company, for the United States Department of Energy’s National Nuclear Security Administration under contract DE-AC04-94AL85000.
4:00 PM - T12: IMF2
BREAK
T13: Fuels II
Session Chairs
Wednesday PM, November 28, 2007
Gardner (Sheraton)
4:30 PM - **T13.1
Molten Salts for Nuclear Reactor Applications.
Ondrej Benes 1 , Rudy Konings 1
1 , European Commission, JRC, Institute for Transuranium Elements, Karlsruhe Germany
Show AbstractThe molten salt reactor is one of the six Generation IV reactor concepts that are being studied at present. The MSR concept is not new and was already explored extensively in the 1960s and early 1970s. At that time the goal was a thermal breeder reactor using 7LiF-BeF2-ThF4-UF4 liquid fuel, producing U-233 from Th-232. This could be achieved by selecting a fuel matrix (7LiF-BeF2) that is almost transparent to neutrons, and applying on-line clean-up of the fuel to remove neutron poisonous fission products. The MSR studied in the frame of Generation IV has the same goal, but the research concerns not only thermal but also epithermal and fast concepts nowadays. In addition, a molten salt reactor concept for transmutation of actinides is studied. These concepts generally require new fuel salts that meet the specific requirements of the various applications. For example, for the molten salt transmutation reactor the solubility of the actinide trifluorides in the matrix is of prime importance, of course in combination with appropriate thermal properties. This renewed attention for the molten salt reactor has also initiated interest in the applications of molten salts as coolant or heat transfer fluids in nuclear reactors, since it is recognised that their heat transfer properties are extremely good. This wide variety of nuclear reactor applications requires a thorough knowledge of the thermodynamic and thermophysical properties of the relevant molten salts. In the design phase the melting point, or more specifically the margin between liquidus temperature and operating temperature, is a key issue. We will discuss how the optimisation of the composition of fuel and coolant salts can be made by thermodynamic assessment, modelling and prediction of binary, ternary and higher order phase diagrams. Various examples will be discussed. Furthermore we will address other relevant physico-chemical properties of the molten salts, such as density, heat capacity and viscosity, aiming to understand eventual non-ideal behaviour and correlate it to excess thermodynamic quantities derived from the modelling.
5:00 PM - T13.2
He Bubbles and Thermal Conductivity of Nuclear Fuel Materials.
Shenyang Hu 1 , Marius Stan 2 , Michael Baskes 2
1 , Pacific Northwest National Laboratories, Richland, Washington, United States, 2 , Los Alamos National Laboratory, Albuquerque, New Mexico, United States
Show Abstract5:15 PM - T13.3
Thermal Properties of Simulated High Burn up Nitride Fuels and Nitride ADS Targets.
Masayoshi Uno 1 , Ken Kurosaki 1 , Shinsuke Yamanaka 1 , Kazuo Minato 2
1 Division of sustainable energy and environmental engineering, Osaka University, Suita, Osaka, Japan, 2 , Japan Atomic Energy Agency, Tokai-mura, Ibaragi, Japan
Show AbstractNitride fuels are under consideration as the advanced fuel of FBR and the targets of Accelerate Driven System (ADS) because of their superior thermal, neutronic properties and so on. However, there is no data of the properties of nitride fuels at high burn-up and ADS targets. In the present study, we made various nitride fuels containing simulated FP elements and nitride targets with inert matrix and evaluated the effect of these additives on the properties of the nitrides.Powder sample of UN was prepared by the carbothermic reduction of uranium dioxide (UO2). NdN, Pd and Mo as FP elements were added to UN powder sample. The amount of the added FP element was determined so that the simulated burn-up was to be 50 and 200 GWd/tU. We also prepared (U0.4Zr0.6)N solid solution and UN+TiN(Ti/(U+Ti)=0.4) compound as the targets for ADS. The thermal expansion, heat capacity and thermal diffusivity of the samples were measured using the thermal dilatometer, DSC and laser-flash method, respectively and the thermal conductivity was estimated.For (U,Nd)N lattice parameter increased with Nd content, thermal expansion did not change and thermal conductivity decreased with Nd content. The thermal expansion and thermal conductivity for Pd contained UN, where Pd precipitated as UPd3 in the grain boundaries of UN, was slightly smaller than those of UN. For Mo contained UN Mo precipitated as Mo metal isotropically. Both the thermal expansion coefficient and thermal conductivity did not vary with Mo content This might result from the low Mo content at this simulated burnup. The thermal conductivity of Mo contained (U,Nd)N decreased according to Nd content.Thermal conductivity of UN in TiN matrix was higher than that of UN but that of UN-ZrN solid solution was smaller than that of UN at lower temperature. This study was carried out within the task “Technological development of a nuclear fuel cycle based on nitride fuel and pyrochemical reprocessing” entrusted from the Ministry of Education, Culture, Sports, Science and Technology of Japan.
5:30 PM - T13.4
Preparation of Nitride Fuel by Spark plasma Sintering Technique.
Hiroaki Muta 1 , Ken Kurosaki 1 , Masayoshi Uno 1 , Shinsuke Yamanaka 1
1 Division of Sustainable Energy and Environemtal Engineering, Osaka University, Osaka Japan
Show AbstractNitride fuel (U, Pu, MA)N (MA=Minor Actinide) has been considered to be one of the candidates of fuel in fast breeder reactor or accelerator driven reactor system. AmN and some minor actinides easily evaporate at high temperatures, hence the sintering temperature and time should be reduced in fabrication process of the nitride fuel pellets. However, it is difficult to fabricate high density nitride pellet due to the high melting point and high chemical reactivity of nitrides. We prepared some nitride pellets using a Spark Plasma Sintering (SPS) technique. The powders of UN, NdN, and DyN were sintered using graphite die under vacuum. High density pellets could be obtained at moderate sintering temperature, under 1500 degree C for all the samples. The sintering finished within 30 minutes, hence the evaporation of minor actinides may be neglected in the process. Additionally the pretreatment of powders, such as addition of sintering agent and ball milling, is not necessary. The SPS process is expected to be suitable as fabrication process of the nitride fuel.
5:45 PM - T13.5
Synthesis and Characterization of Minor Actinide Bearing Nitride Transmutation Fuels.
Christopher Stanek 1 , Stewart Voit 2 , Ken McClellan 1 , Stuart Maloy 1 , John Dunwoody 2 , Robert Margevicius 3 , Thomas Hartmann 4
1 Material Science and Technology, Los Alamos National Laboratory, Los Alamos, New Mexico, United States, 2 Plutonium Manufacturing and Technology, Los Alamos National Laboratory, Los Alamos, New Mexico, United States, 3 PADWP, Los Alamos National Laboratory, Los Alamos, New Mexico, United States, 4 Institute of Nuclear Science and Engineering, Idaho State University, Idaho Falls, Idaho, United States
Show AbstractThe U.S. Department of Energy (DOE) is currently investigating methods to close the nuclear fuel cycle, which by definition involves the reprocessing of spent fuel. The management of spent fuel poses a significant hurdle to the growth of nuclear energy in the United States and worldwide. Near term solutions to reduce the radiotoxic inventory include designing proliferation resistant fuels containing Pu and minor actinides for use in existing light water reactors. However, more significant reductions of the radiotoxic inventory are possible if fast spectrum reactors or accelerator driven systems are considered. Actinide nitrides are a candidate transmutation fuel for these applications, due to significant radiation tolerance, high thermal conductivity and high melting point - all important characteristics of high-burnup fuels. Furthermore, actinide nitrides are mutually soluble.In this presentation, we discuss the fabrication of two categories of minor actinide bearing nitride fuels, namely containing uranium (so-called “fertile”) and uranium-free (“non-fertile”) for the FUTURIX-FTA experiment in the Phenix reactor. Ideally, a transmutation fuel will not contain uranium in order to prevent breeding of plutonium. However, as will be discussed, the uranium free fuel is more difficult to synthesize and may suffer from fuel performance limitations. Similar compositions were also prepared for the AFC-1AE and AFC-1G irradiations in the Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL). In particular we discuss details of the carbothermic reduction - nitridization process including minimization of carbon and oxygen impurities and the retention of minor actinides. Furthermore, we discuss the sintering methods employed to meet target theoretical density values, while avoiding minor actinide volatilization. Finally, we discuss corresponding material characterization of the synthesized pellets. The aim of this presentation is to communicate valuable lessons learned during the synthesis of development transmutation nitride fuels, as well as to highlight questions that remain unanswered.