Symposium Organizers
Chu Chun Fu CEA-Saclay
Akihiko Kimura Kyoto University
Maria Samaras Paul Scherrer Institute
Magdalena Serrano de Caro Lawrence Livermore National Laboratory
Roger E. Stoller Oak Ridge National Laboratory
R1: ODS
Session Chairs
Magdalena Serrano de Caro
Frederic Soisson
Tuesday PM, December 02, 2008
Independence W (Sheraton)
9:30 AM - **R1.1
Nanoclusters in BCC Fe and FCC Ni Alloys: An Experimental and Theoretical Comparison.
David Hoelzer 1 , Chong Long Fu 1 , Xing-Qiu Chen 1 , Jim Bentley 1 , Michael Miller 1
1 , Oak Ridge National Laboratory, Oak Ridge, Tennessee, United States
Show Abstract10:00 AM - R1.2
Phase Stability in Oxide Dispersion Strengthened Steels under In-situ Ion Irradiation.
Djamel Kaoumi 1 , Arthur Motta 1 , Mark Kirk 2 , Bernd Kabius 2
1 Mechanical and Nuclear Engineering , The Pennsylvania State University, University Park, Pennsylvania, United States, 2 Materials and Science Division, Argonne National Laboratory, Argonne , Illinois, United States
Show AbstractOxide dispersion strengthened (ODS) Ferritic/Martensitic steels produced by mechanical alloying with Y2O3 powders are considered possible cladding materials for GEN-IV nuclear reactors (especially fast neutron reactors). They are expected to achieve high creep strength and be resistant to radiation swelling at the temperatures (350-700 C) and high level of doses (~200 dpa) expected in these reactors thanks to the incorporation of a very high density of the fine-scale oxide particles in the matrix. Since these alloys derive their good properties from their special microstructure (especially the oxide dispersion), it is necessary to assess the stability under irradiation of the microstructure in general and the nano-sized particles in particular. The use of in-situ ion irradiation in a TEM provides a unique approach to follow the microstructure evolution of the material under irradiation. Four F/M ODS steels were obtained and characterized using synchrotron XRD in order to detect the oxides which have a small volume fraction: none of the alloys revealed the presence of Y2O3 initially introduced. They were then irradiated in situ in a TEM at ANL both at 25°C and 500°C to doese in excess of 150 dpa. While at 25°C, bigger particles underwent amorphization and/or interfact degradation, at 500°C they appeared to be more stable. Ex-situ EFTEM was used to follow the smaller particles (2-5 nm) at the high doses achieved.
10:15 AM - R1.3
Mechanism of Oxide Particle Evolution in MA957.
Hideo Sakasegawa 1 , Laurent Chaffron 2 , Fabrice Legendre 1 , Yann de Carlan 2 , Loïc Boulanger 1 , Mathilde Brocq 1 , Théodore Cozzika 2 , Joël Malaplate 2
1 DEN/DANS/DMN/SRMP, CEA, Gif sur Yvette France, 2 DEN/DANS/DMN/SRMA, CEA, Gif sur Yvette France
Show Abstract ODS (Oxide Dispersion Strengthened) alloys have superior creep properties. As it is well known, these excellent creep properties result from very fine oxide particles dispersed in the matrix. However, there is no common understanding about the nature of the very small oxide particles. In literatures, two hypotheses about the identification can be seen, 1: non-stoichiometric Y-, Ti-, O enriched clusters and 2: stoichiometric Y2Ti2O7. In our past work, chemically extracted residue method and extraction replica method were applied to the commercial ODS alloy, MA957, and then observations using XRD (X-Ray Diffractometer) and FEG-STEM (Field Emission Gun - Scanning Transmission Electron Microscope) with EDS system (Energy Dispersive X-Ray Spectrometer), respectively. From these results, it is concluded that the composition of small particle is related to the particle size. They had two types of phase at least, 1: non-stoichiometric Y-, Ti-, O-enriched clusters from ~2 to ~15 nm (Y/Ti < 1) and 2: stoichiometric Y2Ti2O7 from ~15 nm to ~35 nm. Based on this result, mechanism of oxide particle evolution is discussed in this presentation, including the latest results regarding detailed observations on chemical compositions and structures.
R2: Structural Materials
Session Chairs
Yasuyoshi Nagai
Yuri Osetsky
Tuesday PM, December 02, 2008
Independence W (Sheraton)
11:00 AM - R2.1
Study of ODS Process Manufacturing by Small Angle Neutron Scattering.
Mathieu Ratti 1 , Didier Leuvrey 1 , Marie-Hélène Mathon 2 , Eric Derniaux 1 , Yann de Carlan 1
1 Service de Recherches Métallurgiques Appliquées, CEA Saclay, Gif sur Yvette France, 2 Laboratoire Léon Brillouin, CEA Saclay, Gif sur Yvette France
Show AbstractThe characterization of ODS process manufacturing, and the influence of titanium on it, has been investigated by mean of the Small Angle Neutron Scattering (SANS) technique. This technique is able to highlight nanometer sized structures present within matrix at each stage of the elaboration of an ODS material:● During the mechanical milling of an ODS powder.● During heat treatments, performed to simulate consolidation process and microstructure evolutions, especially the precipitation of nanometric precipitates within the matrix.A generic study comparing experimental ODS materials with a semi-industrial one was performed. The two experimental materials were obtained from two pre-alloyed ferritic Fe-18Cr-1W (wt %) powders, one with titanium and one without titanium. The semi-industrial one was atomized by Aubert & Duval, mechanical milled with yttria by Plansee and extruded by CEA.Regarding the experimental ODS materials, the SANS study has been extremely interesting and has enabled to clarify mechanisms of dissolution and precipitation of nano-oxides occurring respectively during mechanical milling and hot compaction process. Completed by analyses carried out by SANS, TEM and electronic microprobe on the industrial ODS one, this study has given the opportunity to understand the metallurgy of ODS materials and the control of the precipitation with milling parameters and chemical composition.
11:15 AM - R2.2
Study by X-Ray and Neutron Diffraction of the Precipitation Mechanisms of Oxides in Ferritic ODS Materials.
Eric Derniaux 1 , Mathieu Ratti 1 , Fabien Onimus 1 , Jacques Pele 1 , Hideo Sakasegawa 2 , Yann de Carlan 1
1 Service de Recherches Métallurgiques Appliquées, CEA Saclay, Gif sur Yvette France, 2 Service de Recherches en Métallurgie Physique, CEA Saclay, Gif sur Yvette France
Show Abstract11:30 AM - R2.3
Ion Irradiation Effects on the Mechanical Properties of Nuclear Materials – Micro-bending Test.
M. Pouchon 1 , R. Ghisleni 2 , J. Krbanjevic 3 , J. Chen 1 , J. Michler 2 , W. Hoffelner 1
1 Laboratory for nuclear materials, Paul Scherrer Institute, Villigen PSI Switzerland, 2 , Swiss Federal Laboratories for Materials Testing and Research, Thun Switzerland, 3 , Plasma Physics Research Centre, EPFL Fusion Technology, Villigen PSI Switzerland
Show AbstractTuesday, 12/2New Presenter R2.3 @ 10:30 AMIon Irradiation Effects on the Mechanical Properties of Nuclear Materials – Micro-bending Test. Rudy Ghisleni
11:45 AM - R2.4
Metallurgical Evaluation of High Chromium ODS Steel.
Bassem El-Dasher 1 , James Ferreira 1 , Joseph Farmer 1
1 , Lawrence Livermore National Laboratory, Livermore, California, United States
Show AbstractFerritic-Martensitic ODS steel is one of the candidate materials in fusion reactor cladding design. This is due to a combination of its reduced activation and high temperature strength and creep resistance. Typical compositions of ODS steels contain up to 13% of chromium and up to 1% of yttria, with some alloys containing a small amount of titanium, tungsten and other elements. In this work, we present our assessment of a high (24%) chromium, 1% yttria ODS steel, including metallurgical and corrosion evaluation of as-extruded, processed, and welded specimens. While the motivation of utilizing an ODS steel with a higher chromium content is the potential for improved corrosion performance at high temperatures, the primary drawback is generally considered to be the increased stability of the sigma phase (that is known to negatively affect mechanical properties). At elevated operating temperatures (>700C) however, this is not an issue, as the sigma phase becomes unstable and dissolves back into the matrix.In this presentation, we will report on the microstructural evolution of this high chromium ODS steel as a function of processing and welding parameters and conditions. Characterization results obtained using electron backscatter diffraction and transmission electron microscopy will be presented, with particular attention directed toward features such as the presence and stability of the yttria and sigma phases, as well as grain morphology and texture.This work performed under the auspices of the U.S. Department of Energy by Lawrence Livermore National Laboratory under Contract DE-AC52-07NA27344.
12:00 PM - R2.5
Modeling of the Constitutive Behavior of RAFM Steels under Neutron Irradiation.
Jarir Aktaa 1 , Claus Petersen 1
1 Institut for Materials Research II, Forschungszentrum Karlsruhe, Eggenstein-Leopoldshafen Germany
Show AbstractReduced activation ferritic martensitic (RAFM) steels as structural materials of future fusion power plants will be subjected to complex thermo-mechanical loading and high irradiation doses. Correct modeling of their deterioration under these loading conditions is a precondition of a sufficiently reliable lifetime prediction procedure. Therefore a coupled deformation damage model taking into account the complex non-saturating cyclic softening of RAFM steels has been developed and successfully applied to describe the creep-fatigue behavior of the RAFM steels F82H mod and EUROFER 97 under isothermal cyclic loadings.The coupled deformation damage model developed so far for the reference un-irradiated state has been modified taking into account the influence of irradiation. The modification has been done mainly by adding irradiation hardening variable with an appropriate evolution equation including irradiation dose driven terms as well as inelastic deformation and thermal recovery terms. With this approach the majority of the material and temperature dependent model parameters are no longer dependent on the irradiation dose and only few parameters need to be determined by applying the model to the material behavior in the irradiated state. The modified model has been then applied to describe the behavior of EUROFER 97 observed in the post irradiation examinations of the irradiation programs ARBOR 1 and SPICE. Thereby fairly good results have been obtained in describing the irradiation induced hardening due to neutron irradiation as well as its alteration under inelastic deformation and high temperature dwell conditions.
12:15 PM - R2.6
High Temperature Creep-Fatigue Design.
Farhad Tavassoli 1 , Benjamin Fournier 1 , Maxime Sauzay 1
1 DMN/Dir, CEA, Gif sur Yvette France
Show Abstract12:30 PM - **R2.7
Hybrid Reactor Systems: Challenges and Opportunities.
Tomas Diaz de la Rubia 1
1 Chemistry, Materials, Earth & Life Sciences, Lawrence Livermore National Laboratory, Livermore, California, United States
Show AbstractHybrid reactor systems with fusion neutron sources and sub-critical fission blankets may be able to provide much needed nuclear energy without the need for isotopic enrichment and chemical separation. Such systems will require advanced reactor materials capable of maintaining mechanical integrity, while withstanding extreme temperature and neutron dose. The design of such systems will require materials models adequate to reliably predict component failure under the extreme conditions anticipated in these hybrid reactors, with the calibration and validation of models with a broad range of accelerated testing. This paper will outline related challenges and opportunities for materials science.
R3: Defects
Session Chairs
Rudy Ghisleni
Maylise Nastar
Tuesday PM, December 02, 2008
Independence W (Sheraton)
3:00 PM - **R3.1
In-situ TEM Observation of the One-dimensional Diffusion of Nanometer-sized Interstitial-type Dislocation Loops.
Kazuto Arakawa 1 , Hirotaro Mori 1
1 Research Center for Ultra-High Voltage Electron Microscopy, Osaka University, Ibaraki, Osaka, Japan
Show Abstract3:30 PM - R3.2
Thermodynamics of Interfaces and Surfaces in Fe-Cr from Monte-Carlo Simulations in the Variance Controlled Semi-Grand Canonical Ensemble.
Paul Erhart 1 , Babak Sadigh 1 , Enrique Martinez 1 , Alfredo Caro 1 , Magdalena Caro 1
1 Chemistry, Materials, Earth and Life Sciences Directorate, Lawrence Livermore National Laboratory, Livermore, California, United States
Show AbstractWe have derived a new statistical ensemble for multi-component systems which imposes constraints not only on the average but also the variance of the concentrations. This "variance controlled semi-grand canonical" (VCSGC) ensemble can be readily implemented in Monte-Carlo (MC) simulations and can be parallelized extremely efficiently. MC simulations in the VCSGC ensemble allow us to vary the concentration in a continuous and controlled manner over the entire range and at the same time provide information about the derivatives of the free energy. Unlike the conventional semi-grand canonical ensembles this new method is also applicable to immiscible systems. We have applied this approach in large-scale simulations of interfaces and surfaces in Fe-Cr which demonstrates the parallel efficiency of the method and illustrates the type of thermodynamic information which we can now access.
3:45 PM - R3.3
Effects of Carbon on Defect Properties in bcc Fe From First Principles.
Chu Chun Fu 1 , Estelle Meslin 1 , Christophe Ortiz 2 , Francois Willaime 1 , Maria Caturla 2
1 SRMP, CEA-Saclay, 91191 Gif sur Yvette France, 2 Departamento de Fisica Aplicada, Universidad de Alicante, 03690 San Vicente del Raspeig Spain
Show AbstractSteels play central roles in nuclear technology, in particular as structural materials for fission and future fusion nuclear reactors. Although their mechanical properties have been extensively investigated, relatively little is known about the atomic origin of their macroscopic behavior. First principles calculations provide such information, which is often not directly accessible through experiments. However, Density Functional Theory (DFT) studies in this field is rather new compared with other Material Science disciplines. We report here key DFT contributions that allow to revisit some long-standing questions about effects of C in iron:Carbon --an essential component in steels-- is always present as an interstitial impurity in iron even in ultra-high purity samples. We show that the formation of small C-vacancy complexes are indeed energetically favourable. The corresponding energy gain is mainly dictated by thestrength of the respective C-C covalent bonds. Vacancy diffusivity is shown to be significantly modified by the formation of these complexes exhibiting non-arrhenius behaviour. Effective vacancy diffusivity is calculated as a function of temperature and carbon and vacancy concentrations using thermodynamic and kinetic models. Also, nature and strength of the interaction between C and self-interstitial atoms --elementary radiation defects-- will be presented in detail. These results are discussed in the light of various experimental data.
R4: Fundamental Aspects
Session Chairs
Joseph Farmer
Chu Chun Fu
Tuesday PM, December 02, 2008
Independence W (Sheraton)
4:30 PM - **R4.1
Point Defect and Atomic Flux Couplings in Concentrated Alloys under Irradiation: A Self-consistent Mean Field Approach.
Maylise Nastar 1 , Vincent Barbe 1
1 , CEA Saclay, Gif-sur-Yvette France
Show AbstractWhen an alloy is irradiated, atomic transport can occur through the two types of defects which are created: vacancies and interstitials. Recent developments of the self-consistent mean field (SCMF) kinetic theory could treat within the same formalism diffusion due to vacancies and interstitials in a multi-component alloy. It starts from a microscopic model of the atomic transport and yields the fluxes with a complete Onsager matrix of the phenomenological coefficients. The jump frequencies depend on the local environment trough a 'broken bond model' such that the large range of frequencies involved in concentrated alloys is produced by a small number of thermodynamic and kinetic parameters. Kinetic correlations are accounted for through a set of time-dependent effective interactions within a non-equilibrium distribution function of the system. The Onsager matrix and the atomic fluxes induced by a sinusoidal perturbation are in good agreement with direct Monte Carlo simulations performed in body centred cubic alloys. Using the same mean field approach we study the early stages of a spinodal decomposition under irradiation. In addition to the thermally activated jump frequencies of vacancies and interstitials we consider athermal events such as ballistic atomic exchanges and recombination of vacancies and interstitials within a radius of a few lattice parameters.
5:00 PM - R4.2
Energy Landscape of Small Defect Clusters in Iron.
Mihai-Cosmin Marinica 1 , Francois Willaime 1
1 DEN/DMN Service de Recherches de Métallurgie Physique, CEA Saclay, Gif s/ Yvette France
Show AbstractPredicting the microstructural evolution of the radiation damage in materials requires handling the physics of infrequent-event systems where several time-scales are involved [1]: the diffusion of point defects (self-interstitial atoms, vacancies), the defect clustering (growth or dissociation), the dislocation glide etc. Molecular dynamics simulations are limited in time to typically a few nanoseconds. The other point of view is based on the image at 0 K of the complete energy landscape of the system: the minima configurations and the saddle points which link the local minima. Knowledge of the distribution and properties of these local states can determine the thermodynamics of the systems. The energy landscape of self-interstitial atoms, with a competition between configurations with high and low mobilities is an important issue in the microstructural evolution of ferritic materials (e.g. defect clustering under/or after irradiation)[2]. The interstitial defects in iron have atypical properties and their dynamics are not completely understood [3]. An eigenvector following method for systematic search of saddle points and transition pathways on a given potential energy surface, the activation relaxation technique nouveau [4] (ARTn), is developed, implemented and applied to the physics of small defects in iron. We propose a variation of this method aiming at improving the efficiency of the local convergence close to the saddle point. We prove the convergence and robustness of this new algorithm [5]. The energy landscape at zero K of small interstitial and vacancy clusters (from 1 to 4 defects) is presented using a recent empirical potential. The effect of finite temperature is taken into account using the lattice dynamics. When it is possible the results are discussed and compared to the molecular dynamics simulations performed at higher temperatures. New self-interstitial configurations and migration pathways are reported [6]. [1] C.-C. Fu, J. Dalla Torre, F. Willaime, J.-L. Bocquet, A. Barbu, Nature Mat. 4, 68 (2005)[2] M.-C. Marinica, F. Willaime Solid State Phen. 129, 67 (2007)[3] D.A Terentyev, T.P.C Klaver, P. Olsson, M.-C. Marinica, F. Willaime, C. Domain, L. Malerba, Phys. Rev. Lett. 100, 145503 (2008)[4] G.T. Barkema, N. Mousseau, Phys. Rev. Lett. 77, 4358 (1995)[5] Some improvements of the ART method for finding transition pathways on potential energy surfaces, preprint, E. Cancès, F. Legoll, M.-C. Marinica, K. Minoukadeh, F. Willaime (2008)[6] Energy landscape of small interstitial clusters in iron, preprint, M.-C. Marinica, F. Willaime, N. Mousseau (2008)
5:15 PM - R4.3
The Importance of Spin-orbit Coupling for the Magnetism in Fe and Fe-Cr Alloys.
Roberto Iglesias 2 1 , Maria Samaras 1 , Annick Froideval 1 , Maximo Victoria 1 3 4 , Wolfgang Hoffelner 1
2 , University of Oviedo, Oviedo Spain, 1 , Paul Scherrer Institiute, Villigen Switzerland, 3 , Lawrence Livermore National Laboratory, Livermore, California, United States, 4 , Polytechnic University of Madrid, Madrid Spain
Show AbstractA higher confidence in life-time assessments and design of new materials requires understanding the relevant physical phenomena on a range of scales from the microscopic level of single defect damage effects all the way up to macroscopic effects. To obtain an accurate depiction of defect evolution it is necessary to include mechanisms such as magnetism when studying ferritic materials. In the Fe matrix, the inclusion of Cr leads to anisotropy due to the spin-orbit coupling. Thus employing electronic structure calculations using ab initio methods is necessary to investigate and understand the nature of the interactions present to comprehend the structure of materials.Using FLAPW (Full Potential Linearized Augmented Plane Wave) ab initio calculations, the magnetic properties of the Fe and Fe-Cr system are calculated. The results reveal the strong influence of the different possible local atomic environments. One-to-one comparison with beamline techniques verify the necessity of performing these calculations and show the important role of properly including magnetic effects at a first principles level. Once the important mechanisms have been included at this level, the results can be used as input to form empirical potentials for larger scale simulations.
5:30 PM - R4.4
Electronic Effects in Radiation Damage Simulations.
Dorothy Duffy 1 2 , Alexis Rutherford 1 , Sascha Khakshouri 1
1 London Centre for Nanotechnology, UCL , London United Kingdom, 2 EURATOM/UKAEA Fusion Association, Culham Science Centre, Culham United Kingdom
Show AbstractThe role of electrons in classical molecular dynamics (MD)simulations is generally confined to the interatomic potentials, however in some situations we can expect the electrons to play a more active role. For high energy radiation events, inelastic scattering of electrons results in additional energy loss. Even for low energy radiation the poor representation of thermal conductivity in metallic materials will underestimate the cascade cooling rate, particularly in materials with strong electron-phonon coupling.We have developed a methodology in which electronic effects are included by coupling the atomistic simulation cell to a continuum model for the electronic energy [1]. Electronic thermal diffusion is modeled using a finite difference solution of the heat diffusion equation and energy is exchanged between the lattice and the elctrons at each MD step. We demonstrate that electronic effects can either enhance or decrease the degree of damage, depending on the strength of the electron-phonon coupling [2]. For high energy heavy ion irradiation the energy deposited in the electronic system, by inelastic scattering, may be sufficient to cause local melting and leave a track of defects [3]. In fusion power plants, structural materials will experience irradiation with 14 MeV neutrons, resulting in a spectrum of primary knock-on atom energies with a mean value of around 500 keV in Fe. Such high energy radiation events will be associated with significant electronic excitation and they are beyond the scope of MD cascade simulations. We demonstrate how we can use our coupled MD, electronic transport model to simulate radiation events with fusion relevant energies.[1] D.M.Duffy and A.M.Rutherford “Including the effects of electronic stopping and electron ion interactions in radiation damage simulations” J. Phys. Cond. Mat. 19, 016207 (2007)[2] A.M. Rutherford and D.M. Duffy, “The effect of electron-ion interactions on radiation damage simulations” J. Phys: Cond. Matt. 19, 496201 (2007)[2] D.M.Duffy, N.Itoh, A.M.Rutherford and A.M. Stoneham “Making tracks in metals” J. Phys. Cond. Mat. 20, 082201 (2008)
Symposium Organizers
Chu Chun Fu CEA-Saclay
Akihiko Kimura Kyoto University
Maria Samaras Paul Scherrer Institute
Magdalena Serrano de Caro Lawrence Livermore National Laboratory
Roger E. Stoller Oak Ridge National Laboratory
R5: Radiation Damage
Session Chairs
Tomas Diaz de la Rubia
Dorothy Duffy
Stanislav Golubov
Steve Zinkle
Wednesday AM, December 03, 2008
Independence W (Sheraton)
9:30 AM - **R5.1
3D-AP Analysis Coupled with PAS Measurements for Nano-clusters Observation in RPV Steels and Their Model Alloys.
Yasuyoshi Nagai 1 , Takeshi Toyama 1 , Abderrahim Almazouzi 2 , Masayuki Hasegawa 1 , Eric van Walle 2 , Robert Gerard 3
1 , Tohoku University, Oarai Japan, 2 , SCK/CEN, Mol Belgium, 3 , Tractebel Engineering, Brussels Belgium
Show AbstractMicrostructural change by irradiation causes degradation of materials. Embrittlement of nuclear reactor pressure vessel (RPV) steels by neutron irradiation is one of the examples; precipitation of impurity/solute atoms such as Cu, Ni, Mn, Si etc., defect clusters such as interstitial loops, vacancy clusters and vacancy-impurity/solute complexes, and grain boundary segregation of P, As etc. are the candidates of the microstructural change to affect the mechanical properties of the RPV steels. However, we have no ideal tool which can reveal all of the above changes, unfortunately. Each technique has advantages and disadvantages, and thus it is very important to combine several techniques complementarily. In this sense, positron annihilation spectroscopy (PAS) and three-dimensional atom probe (3D-AP) are an excellent combination. 3D-AP can map the elemental distribution with nearly atomic scale resolution and thus the solute nano-precipitates and the grain boundary segregation can be observed. However, 3DAP cannot detect vacancy-type defects. On the other hand, positron is the only probe that is able to detect vacancy-type defects in metals sensitively. In addition, we have found that positron is also a sensitive probe of the solute nano-precipitates, which enables us to discuss the relationship between 3D-AP and PAS results directly. In this presentation, we will report the recent results of both techniques applied to identical specimens including the RPV model alloys and the surveillance test specimens. We show how this combination is useful to reveal the difference of formation kinetics between the solute nano-precipitates and the matrix defects. In addition, we also introduce our effort on developing a new positron annihilation technique for defect/precipitate study, correlation measurements of positron lifetime and Doppler broadening (so-called, Age-Momentum COrrelation (AMOC) method). This technique has a large potential for quantitative analysis of the precipitate density.
10:00 AM - R5.2
Atom-probe Tomography Analysis of Nanoscale Particles in Oxide-dispersion Strengthened Fe-Cr Steels.
Emmanuelle Marquis 1
1 Department of Materials, University of Oxford, Oxford United Kingdom
Show AbstractReduced activation ferritic and ferritic-martensitic steels (RAFMS) are promising structural materials for the first wall and blanket of future fusion reactors. In order to improve the high temperature materials, nanoscale oxide precipitates were added as they provide dislocation pinning points and remain stable up to close to the melting point. Alloys are typically processed by mechanical alloying following by hot-isostatic pressing. Understanding the role of the nanoscale oxide particles on the mechanical properties and radiation resistance in oxide-dispersion strengthened (ODS) alloys requires knowledge about internal structure, chemistry, and interfacial structure of these particles. Atom probe tomography is one possible technique providing chemical information at the atomic scale and, as shown in previous works on ODS alloys [for instance Larson et al., Scripta Materialia 44 (2001) 359-364 and Miller, et al., Intermetallics 13 (2005) 378-392], the analysis of nanoscale oxides by atom probe tomography is complex.It will be shown that evaporation artefacts strongly affect the three-dimensional reconstruction of these precipitates. However, through careful examination and data analysis, it is possible to gain insight into their atomic structures. The analysis of nanoscale particles in three different ODS alloys (a commercial MA957 alloy, a model ODS Fe-12wt.%Cr alloy, and a ODS Eurofer 97 alloy) will be compared highlighting the similarities between the alloys.The author acknowledges R. Odette (UCSB) for valuable discussions and for supplying the MA957 alloy, V. de Castro, T. Leguey, A. Muñoz, M.A. Monge and R. Pareja (Madrid) for the ODS Fe12Cr alloy, and R. Lindau (FZK, Germany) for the ODS Eurofer steel. This work is funded by the Engineering and Physical Sciences Research Council (EPSRC) under grant number EP/077664.
10:15 AM - R5.3
Microstructure Characterization of ODS-RAFM Steels.
Rodrigo Mateus 1 , Patrícia Carvalho 1 2 , José Correia 3 , Horácio Fernandes 1 , Carlos Silva 1 , Luís Alves 4 , Eduardo Alves 4
1 Euratom/IST, Instituto de Plasmas e Fusão Nuclear, Instituto Superior Técnico, Lisbon Portugal, 2 Euratom/IST, Departamento de Engenharia de Materiais, Instituto Superior Técnico, Lisbon Portugal, 3 Departamento de Materiais e Tecnologias de Produção, LNEG, Lisbon Portugal, 4 Instituto Tecnológico e Nuclear, ITN, Sacavém Portugal
Show AbstractWednesday, 12/3New Presenter R5.3 @ 9:15 AMMicrostructure Characterization of ODS-RAFM Steels. Patricia Carvalho
10:30 AM - R5.4
Radiation Damage Theory: Past, Present and Future.
Alexander Barashev 2 , Stanislav Golubov 1
2 Department of Engineering, The University of Liverpool, Liverpool United Kingdom, 1 Materials Science & Technology Division , Oak Ridge National Laboratory, Knoxville, Tennessee, United States
Show AbstractEfforts of many scientists for more than a half of a century have resulted in substantial understanding of the response of various materials to irradiation. The contribution of theory to this process is significant. Consequently, some phenomena have been predicted before their observation: void swelling, radiation-induced segregation and existence of one-dimensional mass transport under high-energy cascade-producing particle bombardment. Development of the NRT standard for a common measure of the irradiation dose, the Standard Rate Theory and its further development, the BEK model, and finally the Production Bias Model for void swelling, etc. have established a framework for analyzing microstructure evolution in different materials. It has to be admitted, however, that theory has not acquired a status allowing it to play a decisive role in creating radiation-resistant materials. Moreover, some theoretical predictions are in contradiction with observations, which indicates that something important has escaped attention. In the present paper, the current theoretical framework and experimental data are analyzed and the reasons for the situation described are discussed. A way of developing a predictive theory is outlined.
10:45 AM - R5.5
Length-scale Effects in Cascade Damage Production in Iron.
Roger Stoller 1 , Paul Kamenski 2 , Yuri Osetskiy 1
1 Materials Science and Technology Division, Oak Ridge National Laboratory, Oak Ridge, Tennessee, United States, 2 Department of Materials Science and Engineering, University of Wisconsin, madison, Wisconsin, United States
Show AbstractMolecular dynamics simulations have been extensively used in recent years to characterize primary radiation damage formation in the form of atomic displacement cascades. Previous work in nickel has indicated that cascade damage production may be significantly different in nanograined materials. A large database of simulations has been accumulated that describe cascade damage production in single crystal iron using a modified version of the interatomic potential developed by Finnis and Sinclair. This same potential has been used to investigate primary damage formation in nanocrystalline iron in order to have a direct comparison with the single crystal database. A statistically significant number of simulations were carried out at cascade energies of 10 keV and 20 keV and temperatures of 100 and 600K to make this comparison. The variation of cascade size with energy along with the varying grain size demonstrate the influence of grain boundaries as a sink for mobile defects during the cascade cooling phase. Substantially fewer defects survive in the nanograined iron. The influence of the cascade on grain boundary structure and boundary migration was also examined.
11:30 AM - **R5.6
Kinetics of Precipitation in Iron Based Alloys: Thermal Ageing and Irradiation Effects.
Frederic Soisson 1 , Emmanuel Clouet 1 , Chu-Chun Fu 1 , Maximilien Levesque 1 , Maylise Nastar 1
1 , CEA Saclay, Gif-sur-Yvette France
Show Abstract12:00 PM - R5.7
Effects of the Collision Cascade Density on Radiation Damage in Ceramics.
Sergei Kucheyev 1 , A. Azarov 2 , P. Karaseov 2 , A. Titov 2
1 , Lawrence Livermore National Laboratory, Livermore, California, United States, 2 , St. Petersburg State Polytechnic University, St. Petersburg Russian Federation
Show AbstractThe volumetric density of atomic displacements in individual collision cascades is a fundamental parameter of ion-beam defect processes in solids. However, due to complexity of the 3D shape of collision cascades, even the definition of such a density is not straightforward. In particular, all the algorithms previously proposed to calculate cascade densities do not take into account the formation of subcascades, which strongly affect the results. Here, we present a new approach that takes into account subcascades. We perform a statistical analysis of a large number of individual collision cascades with a 3D distribution of vacancies calculated by a binary collision code as an input. We provide examples of how our results can explain some peculiarities of the damage buildup in several non-metallic solids, including Si, GaN, and ZnO. For example, for GaN, results show that an increase in the cascade density above a critical value results in a rapid increase in the rates of damage buildup in the crystal bulk and of planar amorphization proceeding from the surface. This threshold behavior suggests an important role of nonlinear energy spikes in the formation of stable implantation disorder in GaN. The methodology developed in this work can be used in future studies to evaluate the fundamental ion-beam damage mechanisms and in particular to differentiate between contributions from dynamic annealing and energy spikes effects. This work was performed under the auspices of the U.S. DOE by LLNL under Contract DE-AC52-07NA27344.
12:15 PM - R5.8
Atomic-Scale Modeling of Dislocation Dynamics in Radiation Damage Environment.
Yury Osetskiy 1 , Roger Stoller 1 , David Bacon 2
1 , ORNL, Oak Ridge, Tennessee, United States, 2 Engineering, University of Liverpool, Liverpool United Kingdom
Show AbstractDynamics of dislocations in realistic environment of existing microstructure defines mechanical properties of crystalline materials. A vast contribution to mechanical properties depends on the direct interaction between dislocation and other defects and depends very much on the particular atomic-scale structure of the both moving dislocation core and obstacle. In this work we review recent progress in large-scale modeling of dislocation dynamics in metals at atomic level by molecular dynamics and statics. Examples are given for both bcc and fcc metals were edge and screw dislocations were interacting with vacancy (loops, voids, stacking fault tetrahedra, etc), self-interstitial clusters and secondary phase precipitates. Attention is paid to interpretation of atomistic results from the point of view of parameterization of continuum models. The latter is vitally necessary for further application in 3-dimensional dislocation dynamics within the multiscale materials modeling approach.Research sponsored by the Division of Materials Sciences and Engineering and the Office of Fusion Energy Sciences, U.S. Department of Energy, under contract DE-AC05-00OR22725 with UT-Battelle, LLC.
12:30 PM - R5.9
Cr Segregation at the FeCr Surface and the Origin of Corrosion Resistance in Ferritic Steels.
Magdalena Serrano de Caro 1 , Brianah Morse 2 , Nosa Egiebor 2 , J. Farmer 1 , Alfredo Caro 1
1 CMELS, Lawrence Livermore National Laboratory, Livermore, California, United States, 2 Chemical Engineering, Tuskegee University, Tuskegee, Alabama, United States
Show AbstractStructural materials in Gen-IV nuclear reactors will face severe conditions of high operating temperatures, high neutron flux exposure, and corrosive environment. Radiation effects and corrosion and chemical compatibility issues are factors that will limit the materials lifetime. Low-chromium (9-12 Cr wt.%) ferritic martensitic (F/M) steels are being considered as possible candidates because they offer good swelling resistance and good mechanical properties under extreme conditions of radiation dose and irradiation temperature. The surface chemistry of FeCr alloys, responsible for the corrosion properties, is complex. It exists today a controversy between equilibrium thermodynmic calculations, which suggest Cr depletion at the surface driven by the higher surface energy of Cr, and experimental data which suggest the oxidation process occurs in two stages, first forming a Fe-rich oxide, followed by a duplex oxide layer, and ending with a Cr-rich oxide. Moreover, it has been shown experimentally that corrosion resistance of F/M steels depends significantly on Cr content, increasing with increasing Cr content and with a threshold around 10% Cr, below which, the alloy behaves as pure Fe.In an attempt to rationalize these two contradicting observations and to understand the physical mechanism behind corrosion resistance in these materials we perform atomistic simulations using our FeCr empirical potential and analyze Cr equilibrium distributions at different compositions and temperatures in single and polycrystalline samples. We analyze the controversy in terms of thermodynamic and kinetic considerations.This work performed under the auspices of the U.S. Department of Energy by Lawrence Livermore National Laboratory under Contract DE-AC52-07NA27344.
12:45 PM - R5.10
First Principles Study of Silicon Defects and Defect Recombinations in Silicon Carbide.
Guido Roma 1 , Ting Liao 2
1 Service de Recherches de Métallurgie Physique, SRMP/DMN/DEN CEA-Saclay, Gif sur Yvette France, 2 High Performance Ceramic Division, Institute of Metal Research, Shenyang China
Show AbstractThe behaviour of silicon carbide under irradiation, in spite of a large variety of experimental and theoretical works, is not fully understood yet. The large fraction of Si-Si and C-C bonds after or during irradiation, especially at low temperature, lacks a quantitative explanation on the basis of atomic scale processes. In general, the evolution of point defects populations plays an important role and has to be understood in order to be able to predict the modifications of many physical properties whose variations under irradiation can be dramatical (one example is the thermal conductivity). Of course point defects properties are crucial also for the electrical behaviour of silicon carbide as a large band gap semiconductor. In this respect, first principles calculations are a useful tool that, in many cases, allows to obtain some key parameters (formation and migration energies) which are not easily available through experiments.We will review some previous work and present some recent results that we obtained on silicon interstitials which helps to explain some controversies concerning silicon interstitials and recombinations of defect pairs, which could play a very important role for defect kinetics. Finally, we will present some results on structures, energies and possible evolutions of silicon di-interstitials (migration, reorientation, association/dissociation). We will then discuss the implications for the evolution of the structure of SiC under irradiation at the nanoscale.
R6: Coatings
Session Chairs
Jarir Aktaa
David Hoelzer
Arakawa Kazuto
Wednesday PM, December 03, 2008
Independence W (Sheraton)
3:00 PM - **R6.1
Prospects for Evolutionary and Revolutionary Development of Structural Materials for Fission and Fusion Energy.
Steven Zinkle 1 , Jeremy Busby 1
1 Materials Science & Technology Division, ORNL, Oak Ridge, Tennessee, United States
Show AbstractStructural materials for fission and fusion energy systems must be designed to function in an extremely hostile operating environment, incorporating intense neutron and gamma radiation fluxes, high temperatures, corrosive coolants, and high mechanical stresses. Development of structural materials is historically a long and costly process, particularly for nuclear energy applications due to the long proof testing period to validate the performance of the material in prototypic environments for appropriate licensing authorities. As a consequence, the materials in today’s nuclear power plants are based on materials that were originally developed about 50 years ago. For the next generations of fission and proposed fusion power, a key question is whether advanced materials based on modern materials science principles can be incorporated so that the power plants can achieve their full potential. Materials science tools such as computational thermodynamics and multiscale radiation damage computational models in conjunction with rapid science-guided experimental validation (nonirradiation and irradiation environments) may offer the potential for a transformational reduction in the time period to develop and qualify structural materials for advanced nuclear energy systems. Validation of the performance of these advanced materials in prototypic operating environments will be a key step to obtain acceptance of these advanced materials by reactor vendors, utilities, and the licensing authorities. Examples of the potential for rapid development of high-performance structural materials will be given, including both evolutionary ingot-based steel metallurgy and niche processing techniques such as powder metallurgy production of oxide dispersion strengthened steels. In general, both improved mechanical properties and superior radiation stability are obtained by creating a uniform high-density dispersion of highly stable nanoscale particles.
3:30 PM - R6.2
Mechanism of Decarburization of Alloy Inconel 617 in Impure Helium Containing CO and CO2 as Impurities.
Deepak Kumar 1 , Gary Was 1 2
1 Materials Science and Engineering, University of Michigan, Ann Arbor, Michigan, United States, 2 Nuclear Engineering And Radiological Sciences, University of Michigan, Ann Arbor, Michigan, United States
Show AbstractHelium used as a coolant in the Very High Temperature Gas-Cooled Reactors (VHTR) contains part-per-million levels of H2, H2O, CO, CO2 and CH4 as impurities. Depending on the impurity concentrations and temperature, alloy 617, which is the leading candidate alloys for the heat exchanger, can undergo oxidation, carburization and decarburization. These corrosion phenomena, particularly, decarburization is of prime concern as it degrades the creep strength of the alloy. An understanding of the decarburization mechanism is important to develop mitigation strategies.In order to determine the mechanism of corrosion of alloy 617, a simplified gas chemistry containing 12 ppm of CO and 1.5 ppm of CO2 in helium was chosen and corrosion experiments were conducted over the temperature range of 850 - 1000οC. Corrosion coupons were exposed up to 750 hours in a controlled-impurity helium flow system, specifically built in order to control the impurity levels throughout the experiment duration. A discharge ionization detector gas chromatograph (DIDGC) was used to analyze the impurity levels continuously both before and after the furnace to understand the various gas/metal interactions occurring at the surface. The microstructure stability and the surface scale were analyzed using SEM/optical microscope. The composition and structure of the surface scale were analyzed using EDS and XRD.The analysis of the gas mixture coupled with the microstructural examination of the corrosion coupons showed that decarburization of alloy 617 in He +12 ppm CO + 1.5 ppm CO2 occurs only above a critical temperature that lies between 900 and 950οC. Below the critical temperature the carbon potential in the helium is higher than that in the alloy resulting into carburization of the sample via consumption of CO. However, above critical temperature the carbon potential in the environment is lower than that in the alloys resulting in decarburization of the sample via a net production of CO gas. Furthermore the investigation showed that the porous surface scale of Cr2O3 that forms below the critical temperature is unstable above the critical temperature and reacts with the carbon in solution in the alloy to produce chromium and CO(g).
3:45 PM - R6.3
Irradiation Assisted Stress Corrosion Cracking of Alloy 690 in 400°C Supercritical Water.
Rongsheng Zhou 1 , Elaine West 1 , Lumin Wang 1 , Gary Was 1
1 Department of Nuclear Engineering and Radiological Sciences, University of Michigan, Ann Arbor, Michigan, United States
Show AbstractAlloy 690 has been selected as one of the candidate structural materials for the supercritical water reactor (SCWR) for Generation IV nuclear reactors. The irradiation assisted stress corrosion cracking behavior of alloy 690 in the SCWR environment must be determined before it can be considered for use in this reactor. This paper focuses on proton irradiation assisted stress corrosion cracking of alloy 690 in 400°C supercritical water.TEM bars and tensile bars of alloy 690 were proton irradiated using the Tandetron tandem accelerator at the University of Michigan. The irradiations were performed at 400°C with 2 MeV protons to doses of 2, 4 and 7 dpa. Irradiated samples exhibited significant amount of radiation-induced hardening. A constant extension rate tensile (CERT) test of three irradiated 690 specimens was conducted in 400°C SCW. The oxygen content of the SCW was maintained at <10 ppb during the test and the conductivity was <0.1 µS/cm. The samples were strained at a rate of 3×10-7s-1. The results of the 400˚C CERT test showed that all three samples were susceptible to intergranular stress corrosion cracking (IGSCC).Results of detailed microstructure characterization by TEM will also be presented for the better understanding of the stress corrosion cracking mechanism of the alloy.
4:30 PM - R6.4
Ab Initio-Based Modeling of Radiation Effects in Ni-Fe-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 & Engineering, University of Wisconsin - Madison, Madison, Wisconsin, United States
Show AbstractA diffusion model for dilute fcc alloys has been constructed using statistical mechanics based models and the rate expressions from transition state theory. These rates have been parameterized by ab initio calculations for the Ni-Cr and Ni-Fe systems and include vibrational and electronic excitation contributions. This approach provides new understanding of the defect-solute interactions and migration mechanisms in these alloys. New discoveries includes: 1) strong Cr-interstitial binding 2) Weak binding of Cr and Fe to vacancies and 3) Cr diffuses faster through the interstitial flux than Ni or Fe. This model and insight has been applied to the study of radiation-induced segregation (RIS) which has been a studied in Fe-Ni-Cr alloys for over 30 years with no definitive conclusion on the dominant mechanisms. While substantial progress has been made in the area of RIS prediction by empirical fitting, the predictive capability has been hindered by the lack of adequate energetic parameters. For the first time, we parameterize a RIS model with ab initio based diffusion coefficients in the Ni-Fe-Cr system.
4:45 PM - R6.5
Influence of Alloy Microstructure on Oxide Growth in HCM12A in Supercritical Water.
Jeremy Bischoff 1 , Arthur Motta 1 , Lizhen Tan 2 , Todd Allen 2
1 Mechanical and Nuclear Engineering, Penn State University, University Park, Pennsylvania, United States, 2 Engineering Physics, University of Wisconsin-Madison, Madison, Wisconsin, United States
Show AbstractThe Supercritical Water Reactor is one of the six Generation IV nuclear power plant designs and was envisioned for its high thermal efficiency and simplified core. This reactor is designed to function at high temperature (between 500°C and 600°C), which makes the supercritical water environment particularly corrosive. Consequently, corrosion resistance is a key requirement for the cladding and structural materials for this design. The primary candidates for the supercritical water reactor are ferritic-martensitic steels such as HCM12A. Samples were corroded at 600°C for three exposure times: 2, 4 and 6 weeks. In addition, a sample coated with a thin layer of yttrium was corroded at 600°C for 4 weeks in order to investigate the influence of this yttrium surface coating on the corrosion resistance. The oxide layer formed on the HCM12A samples was studied using scanning electron microscopy, and microbeam synchrotron X-Ray diffraction and fluorescence, the latter of which allows sub-micron microstructural and elemental analysis of the oxide layer (up to 0.2 μm step size). Both diffraction and fluorescence information are acquired simultaneously for each step. This analysis showed that the oxide formed a three-layer structure: an outer layer containing Fe3O4, an inner layer containing a mixture of Fe3O4 and FeCr2O4, and a diffusion layer containing a mixture of metal grains and oxide precipitates (mainly FeCr2O4). Some diffraction peaks corresponding to Cr2O3 were observed in the diffusion layer and at the inner oxide-diffusion layer interface. Additionally, SEM images show a strong influence of the base metal microstructure on the oxide advancement. In the base metal, the ferritic-martensitic laths outlined by white spots which may correspond to Cr23C6 are observed. The inner layer is jagged and the diffusion layer is non-uniform, both of which appear to be influenced by the base metal microstructure, such as the orientation of the ferritic-martensitic laths. Furthermore, in the diffusion layer, the oxide precipitates preferentially at the ferritic-martensitic laths boundaries. Chromium-rich Cr23C6 carbides were seen in the base metal and are thought to be located at these laths boundaries. This suggests that the presence of these carbides might significantly influence the oxide precipitation. Concerning the influence of the yttrium surface coating on the corrosion resistance of the material, it was found that the outer layer of the yttrium coated sample was significantly less porous than that of the sample without the yttrium layer. Additionally, a white Y2O3-YFeO3 line is observed in the outer layer which corresponds to the original yttrium coating. No pores are observed in the outer layer beyond this line, thus it appears that the yttrium coating is beneficial to the corrosion resistance of material. This is reinforced by the fact that the yttrium coated sample has slightly thinner oxide thicknesses than the non-coated sample.
5:00 PM - R6.6
Modification of Oxide Multilayer Surfaces by Thermal Treatment and Heavy Ion Beam Irradiation Characterized by AFM.
Marilyn Hawley 1 , David Devlin 1 , Cynthia Reichhardt 2 , Kurt Sickafus 1 , Igor Usov 1 , James Valdez 1 , Yongqiang Wang 1
1 Materials Science and Technology Division, Los Alamos National Laboratory, Los Alamos, New Mexico, United States, 2 Theoretical Division, Los Alamos National Laboratory, Los Alamos, New Mexico, United States
Show AbstractNew fuel composite designs consisting of actinide material imbedded in a radiation tolerant matrix are being explored to fill three critical needs: containment of fission products (FP), easy separation of FP from spent nuclear fuel, and greater energy production. To this end, we have fabricated samples of tri-layer metal oxide films, composed of alternating layers of HfO2 (an actinide surrogate) and MgO, an easily dissolve material, were fabricate by physical vapor deposition methods. This composition and structure was chosen as a model system, ideal for a combined experimental characterization and modeling effort. The multilayer thin films were grown on primarily Si(111) substrates both at room temperature and at ~ 450C. The room temperature grown films were post-annealed in a separate system at temperatures ranging from 200 to 1000C. Changes in microstructure and surface RMS roughness at the surface were tracked using atomic force microscopy (AFM). A significant improvement was seen in the crystal quality, grain size uniformity, and surface roughness of the HfO2 films at the higher temperature. Cracking was observed at the surface of multilayer films annealed at above 500C. Pieces of the multilayer film post-annealed at 550C were irradiated with 10 MeV Au3+ ions over a range of fluences. The degree and nature of the modification of the surface structure of each irradiated sample was characterized using AFM topographic and phase imaging techniques to determine local changes in structure and nano mechanical properties as a function of fluence. Dramatic changes in the surface structures were observed at the highest fluences including the formation of circular depressions often containing a central spire. Single-phase films of the two constituent materials were grown to a comparable total thickness and irradiated for comparison. Details of the experiment and results will be presented.
5:15 PM - R6.7
TiN Coating for Advanced Cladding Development.
Jian Gan 1 , Ick-Chan Kim 2 , Haiyan Wang 2 , Yongli Xu 3 , Joshua Shea 4 , Todd Allen 4 , James Cole 1
1 Nuclear Fuels and Materials, Idaho National Laboratory, Idaho Falls, Idaho, United States, 2 Electrical and Computer Engineering, Texas A&M University, College Station, Texas, United States, 3 Materials Research Lab, UES Inc., Dayton, Ohio, United States, 4 , The University of Wisconsin, Madison, Wisconsin, United States
Show AbstractThe global nuclear partnership program (GNEP) calls for advanced fuel cladding capable of tolerating the high dose irradiation associated with high burn-up transmutation fuels. The cladding must be reliable for high temperature applications and be stable with a reduced rate of fuel-cladding chemical interaction (FCCI). FCCI may be a major life-limiting factor of achieving high burn-up levels for GNEP transmutations fuels. It is not known how the presence of fission product will affect the FCCI. Titanium nitride is known as a material of choice for the coating of cutting and grinding tools as protection against wear, erosion and chemical attack. It is also widely used in microelectronic instrumentation as a diffusion barrier and adhesion promoter. The ability of titanium nitride to act as a diffusion barrier combined with its mechanical and thermal properties make it attractive for nuclear applications. This work investigates the feasibility of applying TiN coating at the inner wall of HT-9 cladding. TiN films at thickness of 1.0 μm and 5.0 μm were coated using pulsed laser deposition (PLD) or cathodic arc deposition (CAD). The mechanical properties of the thin coatings were evaluated using micro- and nano-hardness. Proton irradiations were conducted at 500C to 0.5 dpa and 2.0 dpa. The cross-section TEM analysis will be carried out to study the radiation effect in the microstructure of TiN and the TiN/HT-9 interface.
5:30 PM - R6.8
A Surface Modification of Hastelloy X by a SiC Coating and an Ion Beam Irradiation for a Potential use for Iodine-Sulfur Cycle in Nuclear Hydrogen Production System.
Jae-Won Park 1 , Zuhair Khan 1 , Hyung-Jin Kim 1 , Yongwan Kim 1
1 , Korea Atomic Energt Research Institute , Daejeon Korea (the Republic of)
Show AbstractThe materials used for the SO3 decomposer in Iodine-Sulfur (IS) cycle for Nuclear Hydrogen Production System require excellent mechanical properties as well as a high corrosion resistance in SO2/SO3 environment at an elevated temperature up to 950 degree C. So far, no metallic materials have been suggested to be useful in such an environment. A surface modification of Hastelloy X by a SiC coating processed by an electron beam evaporative deposition has been studied in combination with an ion beam mixing (IBM) and an ion beam hammering (IBH). The simply deposited SiC film on the Hastelloy X substrate is easily peeled-off during an annealing at a high temperature due to a huge difference in their thermal expansion coefficients, however the SiC coating on Hastelloy X prepared with IBM is sustained above 900 degree C when the heating rate is less than 10 degree C/min. The process of coating and IBM consists of a thin SiC film deposition, a subsequent N ion beam bombardment, and then an additional deposition of the film to the designed thickness. IBM plays a role of fastening the SiC film on the Hastelloy X substrate until the interfacial reaction takes place. Once the reaction takes place, new phases are developed at the interface under the consumption of the film and the substrate materials, producing a buffer layer. Without IBM, the SiC film tends to be easily detached during an annealing before the interfacial reaction initiates.The SiC film prepared with IBM requires a post-deposition annealing in vacuum for the interfacial reaction. However, the sublimation of SiC film prepared by an electron beam evaporative deposition occurs at the temperature above 900 degree C, decreasing the thickness of the deposited film. The sublimation of the SiC film can be prevented by IBH in which ion beams are bombarded onto the deposited film. This may be attributed to an ion beam bombardment induced densification of the deposited film.The resultant SiC coated Hastelloy X prepared by IBM and IBH exhibits a high corrosion resistance in a sulfuric acid at 300 degree C, suggesting a possible application for the IS cycle.
5:45 PM - R6.9
Corrosion of Ferritic Steels in High Temperature Molten Salt Coolants for Nuclear Applications
Joseph Farmer 1 , James Ferreira 1 , Magdalena Serrano de Caro 1
1 Chemistry, Materials, Earth & Life Sciences, Lawrence Livermore National Laboratory, Livermore, California, United States
Show AbstractCorrosion and environmental cracking of ferritic steels in high temperature molten fluoride salts may limit the life of advanced reactors, including some hybrid systems that are now under consideration. In some cases, the steel may be protected through galvanic coupling with other less noble materials with special neutronic properties such a beryllium. This paper reports the development of a model for predicting corrosion rates for various ferritic steels, with and without oxide dispersion strengthening, in FLiBe (Li2BeF4) and FLiNaK (Li-Na-K-F) coolants at temperatures up to 800 degrees Centigrade. Mixed potential theory is used to account for the protection of steel by beryllium, Tafel kinetics are used to predict rates of dissolution as a function of temperature and potential, and the thinning of the mass-transfer boundary layer with increasing Reynolds number is accounted for with dimensionless correlations. The model also accounts for the deceleration of corrosion as the coolants become saturated with dissolved chromium and iron. This paper also reports electrochemical impedance spectroscopy of steels at their corrosion potentials in high-temperature molten salt environments, with the complex impedance spectra interpreted in terms of the interfacial charge transfer resistance and capacitance, as well as the electrolyte conductivity. Such in situ measurement techniques provide valuable insight into the degradation of materials under realistic conditions.
R7: Poster Session
Session Chairs
Maria Samaras
Roger Stoller
Thursday AM, December 04, 2008
Exhibition Hall D (Hynes)
9:00 PM - R7.1
Chemical and Irradiation Stability of the Spinel Compound Magnesium Stannate (Mg2SnO4).
Peng Xu 1 , Juan Nino 1
1 Materials Science and Engineering, University of Florida, Gainesville, Florida, United States
Show AbstractInert matrix (IM) materials for nuclear fuel in light water reactors (LWRs) must meet several critical requirements that include high temperature stability, good irradiation resistance, high thermal conductivity, low solubility in hot water, and feasible aqueous reprocessing. Due to the excellent irradiation stability against thermal and fast neutrons, normal spinel MgAl2O4 has received significant attention as a potential IM candidate. However, during in-pile testing this compound failed as a result of irradiation damage caused by high energy fission fragments that led to swelling. Mg2SnO4 is an inverse type spinel with similar neutronic properties to MgAl2O4 but may possess better irradiation tolerance against fission fragments, since Sn4+ is a much larger and heavier ion than Al3+, making it less susceptible to irradiation damage. In this work, Mg2SnO4 was synthesized through conventional solid state processing. Its irradiation stability was characterized using in situ TEM technique at Argonne National Lab (ANL). Atomic displacement damage and implantation concentration were calculated using TRIM-2008. The Kr2+ ion source at 1 MeV was used for irradiation and structural evolutions were monitored and recorded through bright field (BF) images and selected area electron diffraction (SAED) patterns. The amorphization of Mg2SnO4 was achieved at an ion dose of 5×1019 Kr ions/m2 at 50 K and 1020 Kr ions/m2 at 150 K, which is equivalent to a peak atomic displacement damage of 6.1 dpa and 12.2 dpa, respectively. The spinel crystal structure was thermally recovered at room temperature from the amorphous phase caused by irradiation at 50 K; however, phase separation was observed near the edge of the specimen. In terms of chemical stability, it is shown that Mg2SnO4 has good corrosion resistance against water at 300 °C, with minimum swelling and solubility (less than 1% in mass and volume change). Further aqueous dissolution tests indicate that all Mg2+ can be completely dissolved in HNO3 and the sintered body can be broken up and powderized by magnetic bar stirring.
9:00 PM - R7.10
Preparation of SiCf/SiC Composites for Fusion Applications using Ceramic Routes.
Goran Drazic 1 , Sasa Novak 1 , Tea Toplisek 1 , Katja Konig 1 , Aljaz Ivekovic 1
1 Nanostructured materials, Jozef Stefan Institute, Ljubljana Slovenia
Show AbstractThe production of SiC/SiC composites with the properties required for fusion application is a complex, multi-stage process. The CVI and PIP techniques result in very low-activation materials but with unacceptable residual porosity. On the other hand, the NITE method produces a dense material with good mechanical properties, but the used sintering additive used can make a considerable undesirable contribution to the activity of the material. In both cases the serious problem is relatively low thermal conductivity of the material.This paper will present an alternative approach to the production of SiC/SiC composite using a ceramic processing route. In the investigation, special attention was given to the selection of low-activation sintering additives that enable densification of the matrix material at moderate temperatures, i.e., below 1500°C. The compositions were tailored with respect to the calculated activation in a fast-neutron flux and taking into account the thermal stability of the available SiC fibres. The technique comprises infiltration of SiC-fiber perform with a colloidal suspension of micron- and nano-sized powders mixture and in the second stage infiltration with sintering aids based on MgO (Al2O3) - SiO2 - P2O5 system. With the aim to minimize the amount of secondary phase in the matrix material, the powders were coated with a thin layer of MgO or Al2O3. Special attention will be focused on infiltration techniques where SiC based cloth is impregnated with ceramic slurry using vacuum slip infiltration or electrophoretic deposition.
9:00 PM - R7.11
Development of an Electrostatic Dust Removal Grid Array Device for the International Thermonuclear Experimental Reactor.
Jane Leisure 1 , Charles Skinner 2 , Sigurd Wagner 1
1 Department of Electrical Engineering, Princeton University, Princeton, New Jersey, United States, 2 , Princeton University, Princeton, New Jersey, United States
Show Abstract9:00 PM - R7.12
Radiation Damage Studies of Oxide Thin Films.
Natalya Suvorova 1 , Igor Usov 1 , Ming Tang 1 , Joshua Williams 1 , James Valdez 1 , Cynthia Olson Reichhardt 1 , Michael Hundley 1 , Kurt Sickafus 1
1 , Los Alamos National Laboratory, Los Alamos, New Mexico, United States
Show Abstract9:00 PM - R7.13
Surface Engineering Techniques for Improvement of Structural Nuclear Materials Properties.
Victor Andrei 1 , Constantin Diaconu 1 , Cristian Lungu 2 , Gheorghe Oncioiu 1 , Manuela Fulger 1 , Catalin Ducu 3
1 Surface Analysis Laboratory, Institute for Nuclear Research, Mioveni Romania, 2 , National Institute for Laser, Plasma and Radiation Physics, Bucharest Romania, 3 , University of Pitesti, Pitesti Romania
Show Abstract9:00 PM - R7.14
Transmission Electron Microscopy Study on Electron Beam Irradiation Induced Phase Transformation of Niobium Nitride.
Jonghan Won 1 , James Valdez 1 , Kurt Sickafus 1 , Manabu Ishimaru 2 , Muneyuki Naito 2
1 , Los Alamos National Laboratory, Los Alamos, New Mexico, United States, 2 , Institute of scientific and industrial research, Ibaraki, Osaka, Japan
Show AbstractThe Nb-N binary system has received considerable attention over recent decades due to the interesting physical properties of compounds in this system, such as the superconducting behavior of the δ-NbN1-x phase. However, the phase transformation behavior and structural evolution of niobium nitride phases under electron irradiation have not been investigated yet. In this study, we examine electron beam irradiation-induced changes in the crystal structure of non-stoichiometric niobium nitride (NbN1-x). Structural changes were monitored using transmission electron microscopy (TEM).For these experiments, a pristine sample of NbN1-x was obtained from a commercially available bulk sputtering target. The initial crystal structure of the pristine sample was determined using grazing incidence X-ray diffraction (GIXRD). GIXRD indicated that the pristine sample possesses a γ-Nb4N3 structure (I4/mmm), whose unit cell is nitrogen vacancy ordering tetragonal superstructure of the cubic δ-NbN (Fm-3m) phase (the unit cell of γ-Nb4N3 is twice the volume of the δ-NbN unit cell). The atomic configuration of γ-Nb4N3 consists of an ordered arrangement of nitrogen vacancies, which are responsible for the superlattice reflections observed in the GIXRD patterns. Also, these nitrogen vacancies induce a ~5 % volume shrinkage of γ-Nb4N3 compared to δ-NbN. After TEM electron beam irradiation to an electron fluence of 5.4×1022 e/cm2 (300 keV electrons), selected-area electron diffraction (SAED) patterns obtained from the irradiated sample volume indicated that the γ-Nb4N3 superlattice reflections completely disappeared. It is assumed here that the mechanism for this reflection extinction involves displacement of N atoms by incident electrons. Specifically, the electron beam irradiation causes nitrogen atoms to jump into the vacant N sites. Eventually, this displacement damage mechanism leads to an order-to-disorder transformation of the NbN1-x compound. The calculated displacement damage dose necessary to achieve this order-to-disorder transformation (in units of displacements per atom or dpa) is ~0.05 dpa (assuming a threshold energy for atomic displacement of N atoms given by Ed = 40 eV). This dpa seems rather small in order to achieve a complete order-to-disorder transformation. Thus, it is likely that the displacement threshold, Ed, for N atoms in non-stoichiometric NbN1-x is much lower than 40 eV.
9:00 PM - R7.15
Microstructural Study of Oxide Scales Formed on Alloy HCM12A in Supercritical Water by TEM.
Pantip Ampornrat 1 , Yanbin Chen 1 , Lumin Wang 1 , Gary Was 1
1 Nuclear engineering and radiological sciences, University of Michigan, Ann Arbor, Michigan, United States
Show Abstract9:00 PM - R7.17
Synthesis and Characterization of Large Area cBN Films by Short-pulse Laser Plasma Deposition Techniques.
Peter Feng 1
1 Physics department, University of puerto rico, San Juan, Puerto Rico, United States
Show Abstract9:00 PM - R7.18
Magnetism of Fe-Cr alloys: An Element-resolved PEEM / XMCD Study.
Annick Froideval 1 , Roberto Iglesias 1 2 , Maria Samaras 1 , Manuel Pouchon 1 , Jiachao Chen 1 , Stefan Schuppler 3 , Peter Nagel 3 , Maximo Victoria 1 4 5 , Wolfgang Hoffelner 1
1 Nuclear Energy and Safety, Paul Scherrer Institut, Villigen Switzerland, 2 Departamento de Física, Universidad de Oviedo, Oviedo Spain, 3 Institut für Festkörperphysik, Forschungszentrum Karlsruhe, Karlsruhe Germany, 4 Chemistry and Materials Science, Lawrence Livermore National Laboratory, Livermore, California, United States, 5 Institute of Nuclear Fusion, Polytechnic University of Madrid, Madrid Spain
Show AbstractIndustrial steels, whose chemical compositions are based on an Fe-Cr matrix with Cr concentrations ([Cr]) ranging from 2-20 atomic percent (at. %), are possible candidates foreseen for the design of structural components in advanced nuclear energy installations such as Generation IV and fusion reactors. Recent theoretical calculations predict a negative sign of the heat of formation of Fe-Cr alloys with [Cr] lower than 10-12 at. % [1, 2] emphasizing the crucial role played by magnetism in the stability of the Fe-Cr system. A complete understanding of the magnetic properties of Fe-Cr alloys is thus crucial to meet the challenge of developing advanced nuclear materials resistant to high radiation doses and temperatures. In this context, element-resolved synchrotron x-ray experiments using photoemission electron microscopy (PEEM), and x-ray magnetic circular dichroism (XMCD), coupled with ab initio calculations have already been performed on two Fe-Cr alloys containing 6.2 and 12.7 at.% Cr [3]. These magnetic investigations have revealed that the Cr content of the Fe-Cr alloys strongly influences the ferromagnetic microstructure and the values of the iron magnetic moments of such binary alloys. In order to obtain a detailed understanding of the magnetic properties of these alloys on Cr concentration, a series of Fex-Cr1-x alloys with x comprised between 0.84 and 0.98 have also been measured using the PEEM / XMCD techniques. The values of the orbital and spin moment ratios of iron in Fe-Cr alloys have been determined from the XMCD spectra according to the sum rules data analysis described by Chen et al. [4]. An interesting interdependence between magnetism and stability of Fe-Cr alloys has been determined. References[1] P. Olsson, I.A. Abrikosov, and J. Wallenius, Phys. Rev. B 73, 104416 (2006).[2] T.P.C. Klaver, R. Drautz, and M.W. Finnis, Phys. Rev. B 74, 094435 (2006).[3] A. Froideval, R. Iglesias, M. Samaras, S. Schuppler, P. Nagel, D. Grolimund, M. Victoria and W. Hoffelner, Phys. Rev. Lett. 99, 237201 (2007).[4] C. T. Chen, Y. U. Idzerda, H.-J. Lin, N. V. Smith, G. Meigs, E. Chaban, G. H. Ho, E. Pellegrin, and F. Sette, Phys. Rev. Lett., 75, 152, 1995.
9:00 PM - R7.19
Stability of the Particle Dispersion of an ODS/Fe12Cr Alloy after Ion Irradiation.
Vanessa de Castro 1 , Sergio Lozano-Perez 1 , Emmanuelle Marquis 1 , Mike Jenkins 1
1 Department of Materials, University of Oxford, Oxford United Kingdom
Show AbstractFerritic/martensitic steels to be used as structural materials in a future fusion reactor have to be resistant to the damage induced by the energetic neutrons that result from the fusion reaction. The maximum service temperature of these materials is established at around 550 C. One way to increase this temperature in 100 C or more is to homogeneously disperse hard nanosized oxide particles, such as Y2O3, into the steel matrix. It has been demonstrated that oxide dispersion strengthened (ODS) steels have better tensile and creep properties than the unreinforced ones [1]. The poor impact behavior achieved in the 1st generation ODS steels significantly improves when applying suitable thermo-mechanical treatments [2]. The particles would also help to lower the rate of damage accumulation in the steel if they acted as trapping sites for irradiation induced defects and He bubbles, inhibiting their growth and migration to grain boundaries [3]. In order to prove the efficiency of these materials it is of great importance to analyze their microstructures after irradiation. The ODS steels are normally produced by mechanical alloying and consolidated by hot isostatic pressing (HIP) or hot extrusion to provide a homogeneous dispersion of the nanoparticles in the matrix. These production routes lead to very complex microstructures were it is difficult to separate the role of the different alloying elements and microstructural features.In the present work a less complex ODS system based on a Fe12Cr binary alloy has been studied after ion irradiation in order to achieve a better understanding of the role of the ODS particles in damage development. The ODS/Fe12Cr alloy and a reference Fe12Cr alloy were irradiated at 500 C with 0.5 and 2 MeV Fe+ ions in order to obtain a flat damage profile within about 1 micron of the ion-entry surface. Alloys irradiated in this way at doses from 1014 to 1016 ions/cm2 have been characterized by transmission electron microscopy and atom probe tomography. The results of these investigations will be presented. The authors acknowledge T. Leguey, A. Muñoz, M. A. Monge and R. Pareja for providing the alloys. This research has been supported by FP6 Euratom Research and Training Programme on Nuclear Energy.[1] R. Schäublin et al., J. Nucl. Mater. 307-311 (2002)778.[2] R. Lindau et al., Fus. Eng. Des. 75-79 (2005) 989.[3] S. Ukai and M. Fujiwara, J. Nucl. Mater. 307-311 (2002) 749.
9:00 PM - R7.2
Radiation Tolerance of MAX Phase Alloys Using In-situ Ion Irradiation Techniques.
Karl Whittle 1 , Daniel Riley 3 , Mark Blackford 1 , Katherine Smith 1 , Nestor Zaluzec 2 , Sam Moricca 1 , Gregory Lumpkin 1
1 Institute of Materials Engineering, ANSTO, Sydney, New South Wales, Australia, 3 School Of Engineering, University of Melbourne, Melbourne, Victoria, Australia, 2 Materials Science Division, Argonne National Laboratory, Chicago, Illinois, United States
Show AbstractRecently it has been proposed that materials based on MAX phase compositions have been proposed as having applications within the future nuclear technologies. In order for a material to have potential applications in fission and fusion technologies the ability of the materials to tolerate radiation damage needs to be understood and in particular predicted.Two systems in particular Ti3AlC2 and Ti3SiC2 have been studied with a view to determining their radiation tolerance, using in-situ ion beam irradiation of 1MeV Kr ions, coupled with transmission electron microscopy. Experiments have shown that Ti3AlC2 in particular showing no appreciable damage at 300K up to doses of at least 3.75x1015 ions cm-2 (~12 dpa). Interestingly there is an observed difference between Ti3AlC2 and Ti3SiC2.Explanations and possible mechanisms for recovery from damage are presented, along with implications for future potential uses.
9:00 PM - R7.20
Ion Irradiation Damage Effects on the Mechanical Response of Ferritic ODS Alloys.
Rudy Ghisleni 1 , Manuel Pouchon 2 , William Mook 1 , Jiachao Chen 2 , Wolfgang Hoffelner 2 , Johann Michler 1
1 , Swiss Federal Laboratories for Materials Testing and Research (EMPA), Thun Switzerland, 2 Laboratory for Nuclear Materials, Paul Scherrer Institute, Villigen-PSI Switzerland
Show AbstractFerritic oxide dispersion strengthened (ODS) steels (like PM2000), are candidate materials for nuclear plants such as Generation IV and fusion reactors due to their good mechanical properties at elevated temperatures. Since these ODS steels are based on iron, less helium embrittlement is expected in comparison to nickel based alloys. The determination of mechanical properties with sub-sized samples is a necessary requirement for the understanding of materials deformation from its very early phase to failure. The response of microstructure to mechanical load is studied by observing the deformation of micropillars mechanically loaded with a microindenter head inside a SEM. The microcompression test allows a direct determination of stress-strain curves.This study is centered on the investigation of the mechanical properties degradation as a function of irradiation exposure. The elastic modulus and yield stress are evaluated by in situ (SEM) compression of micropillars fabricated by focused ion beam. The micropillars diameter is varied from 300 nm to 1 μm. The irradiations were performed at room temperature with 1.5 MeV 4He2+ ions at different fluences ranging from 1.4x1016 to 11.2x1016 ions/cm2 with each irradiation being performed at incidence angles ranging from 0 to 66 deg. The average yttria dispersoids size is evaluated to be 28 nm while the grain size is evaluated to be in the mm to cm range by microstructure analysis. The microcompression test results are compared with standard nanoindentation results.
9:00 PM - R7.21
Fracture Toughness Properties of Nanostructured Ferritic Alloys Before and After Irradiation.
Mikhail Sokolov 1 , David Hoelzer 1
1 , ORNL, Oak Ridge, Tennessee, United States
Show Abstract9:00 PM - R7.24
Diffusion Controlled Mobility and Multistring Frenkel-Kontorova Model for Screw Dislocations in Bcc Transition Metals.
Mark Gilbert 1 2 , Sergei Dudarev 1 4 , Peter Derlet 3 , David Pettifor 2 , Jaime Marian 5 , Vasily Bulatov 5
1 , EURATOM/UKAEA Fusion Association, Abingdon United Kingdom, 2 Department of Materials, University of Oxford, Oxford United Kingdom, 4 Department of Physics, Imperial College, London United Kingdom, 3 , Paul Scherrer Institute, Villigen, PSI Switzerland, 5 Chemistry, Materials, Earth, and Life Sciences Directorate, Lawrence Livermore National Laboratory, Livermore, California, United States
Show AbstractBody-centred-cubic (bcc) transition metals, in particular Fe and W, and their alloys are the subject of much interest as candidate materials for the critical components for Generation IV Fission reactors and concept designs for Fusion power plants, where they must be able to withstand significant stresses and temperatures. One of the fundamental factors controlling the ability of a metal to plastically deform in a brittle or a ductile mode is the mobility of screw dislocations. There has been much effort to model the motion of screw dislocations using molecular dynamics, but without exception this has thus far only been achieved in the regimes where the stresses were many orders of magnitude higher than those realised in laboratory experiments. We present results from new large-scale molecular dynamics (MD) simulations designed to measure the thermally activated diffusion of screw dislocation lines occurring at finite temperature in Fe, both in the absence of applied stress and in the limit of small stresses.The governing process behind the motion of screw dislocations is the formation and propagation of double-kinks along the dislocation line. An understanding of this mechanism requires appreciation of the core structure of a screw. In this work we develop a new method for the study of the core of screw dislocations based on the ab-initio parameterised multistring Frenkel-Kontorova model. In this model we consider one-dimensional deformable strings of atoms and construct effective inter-string potentials to describe the interaction of strings using both density-functional data and semi-empirical potentials derived for some of the bcc metals. We find that the structure of the core is completely determined by the interaction of strings with their first-nearest-neighbours, and show how subtle changes in the shape of the inter-string potential can dramatically alter the core structure of a screw dislocation.
9:00 PM - R7.25
Diffusion in Bcc Fe Based Alloys Using First Principles Approach.
Samrat Choudhury 1 , Benjamin Swoboda 1 , Julie Tucker 1 , Dane Morgan 1
1 Materials Science and Engineering, University of Wisconsin, Madison, Madison, Wisconsin, United States
Show AbstractMechanical properties of structural materials used in nuclear reactors are known to deteriorate in irradiation environments. Radiation changes the local composition of the materials through the formation and migration of large concentration of point defects. Hence, analyzing the fundamental mechanism of material transport in irradiation environments is not only critical to understand radiation induced segregation (RIS) but such knowledge of diffusion of chemical species can also be applied in designing multi-component alloy for next generation of reactors. Prior theoretical models to study diffusion in multi-component alloy are often restricted due to lack of availability of thermodynamic and kinetic parameters of the diffusing species. In this work, as a first example, we studied the point defect diffusion of species in ferritic Fe-Ni and Fe-Cr dilute alloy system using first principles. Calculated values will be compared with previous first principles studies and experiment. Diffusion constants for the dilute limit will be determined with the five frequency model. Such study in binary Fe based alloys will provide information for mapping specie dependent diffusion in the whole Fe-Ni-Cr ternary alloy system. Finally, we will discuss the possibility of calculating diffusion constants using the thermokinetic parameters obtained from ab initio calculations in combination with the cluster expansion formalism and Kinetic Monte Carlo (KMC) approach.
9:00 PM - R7.27
Helium Behaviour in Radiation Damaged Zircon (ZrSiO4): A Combination of Classical MD and DFT Study.
Iman Saadoune 1 , Jeremy Rabone 1 , Nora de Leeuw 1
1 Chemistry , University College London, London United Kingdom
Show AbstractZircon (ZrSiO4) is a tetragonal orthosilicate ceramic. Due to its durability, high phase stability and ability to incorporate up to 10% of uranium and plutonium, zircon has been proposed as a host matrix to encapsulate highly radioactive materials from dismantled weapons and nuclear waste from power stations. In these applications, radioactive species such as uranium and plutonium emit α radiation, which causes extensive radiation damage in the system leading to amorphization of parts of the material. After a period of time, the α particles transform into stable He nuclei, that get trapped in the amorphous regions. Insight into the nature of He diffusivity and its interaction with radiation-damaged zones in the zircon lattice is of fundamental importance if we are to advance our understanding of this material and enhance its performance for nuclear disposal applications.Here we present results from a classical molecular dynamics simulation of He diffusion in radiation-damaged zircon, where we apply a novel methodology [1] to create high energy radiation damage zones (up to 15000 kJmol-1/Å) in the zircon system. We also show results from our Density Functional Theory (DFT) calculations addressing the energy pathways of He diffusion in the defective zircon lattice and the electronic nature of He interaction with different defect species that are produced by radiation damage events in the zircon lattice. From our DFT results we find that the presence of defect species; e.g. cation and anion vacancies and their Frenkel pairs, alter greatly the solubility and mobility behaviour of He in the zircon mineral. Anionic oxygen vacancies act as F centres, trapping two electrons where negative charges repel He from the vacancy sites. However, the presence of interstitial oxygen in the lattice reduces the extent of repulsion of He from the oxygen vacancy site, hence lowering significantly the energy barrier for He incorporation site by as much as 54 kJ/mole and trapping it into the vacancy. The presence of an F centre in the close vicinity of He also makes its incorporation in interstitial sites more favourable than in the perfect lattice, which either lowers the energetic barrier for He diffusion between interstitial sites in the high-energy pathways or impedes He diffusivity in the energetically accessible pathways. Our results explain experimental findings regarding the effect of radiation damage on He behaviour in zircon [2].1.J.A.L.Rabone, A.C., A.Hurford and N.H. de Leeuw, Modelling the formation of fission track in apatite using MD simulations. Physics and Chemistry of Minerals, 2008.2.Flowers, R.M., et al., Radiation damage control on apatite (U-Th)/He dates from the Grand Canyon region, Colorado Plateau. Geology, 2007. 35(5): p. 447-450.
9:00 PM - R7.28
Radiation Damage in UO2 by Molecular-Dynamics Simulation.
Dilpuneet Aidhy 1 , Tapan Desai 2 , Paul Millett 2 , Taku Watanabe 1 , Simon Phillpot 1 , Dieter Wolf 2
1 Materials Science and Engineering, University of Florida, Gainesville, Florida, United States, 2 Material Sciences Department, Idaho National Laboratory, Idaho Falls, Idaho, United States
Show AbstractWe illustrate the evolution of the point defects, Frenkel pairs (FPs), during the kinetically-driven phase of irradiation in UO2. During the initial thermal-spike phase, large numbers of point defects on both uranium and oxygen sub-lattices are created. In addition to electronic and microstructural changes, the thermal spike phase is followed by kinetically-driven point defect recombination. The defects that remain after such recombination, however, cause long-term damage to the material. Therefore, it is extremely important to understand the nature of the long-lived defects and defect clusters. In a UO2 system with a high concentration of point defects, modeled by molecular-dynamics (MD) simulation, we have found that the kinetic-evolution of the FPs the two species are strongly coupled. For example, in the absence of the uranium defects the oxygen FPs recombine very quickly. By contrast, when uranium defects are present, the oxygen defects tend to cluster. We further show that under certain circumstances, the lattice itself responds by further contributing to the number of the point defects. This work was supported by DOE NERI contracts DE-FC07-07ID14833 and DE-FC07-05ID14649 and by the DOE-BES Computational Materials Science Network.
9:00 PM - R7.29
Xenon Migration Behavior in UO2.
Jianguo Yu 1 , Ram Devanathan 1 , William Weber 1 , Emily Moore 1 2 , L. Rene Corrales 2
1 , Pacific Northwest National Laboratory, Richland, Washington, United States, 2 Department of Chemistry, University of Arizona, Tucson, Arizona, United States
Show AbstractFission gas release in UO2 is an important issue in nuclear reactors and has attracted a large number of investigations. However, the evolution of fission gas bubbles in UO2 and their interaction with the microstructure are poorly understood. In this work, molecular dynamics simulations and DFT calculations are used to investigate the atomic migration behavior of Xe in UO2. The diffusion properties of Xe are of interest under reactor operating conditions in relation to thermal-directed migration, radiation-assisted migration and the formation of Xe bubbles. The interaction of Xe bubbles with displacement damage will also be presented. The results of this work will be compared to existing experimental data in an attempt to provide an atomistic picture of some of the important factors underlying the diffusion of Xe in UO2.
9:00 PM - R7.3
Influence of Fast Neutron Irradiation on the Mechanical Properties and Microstructure of Nanostructured Metals/Alloys.
Walid Mohamed 1 , K. Murty 1
1 Nuclear Engineering , North Carolina State University, Raleigh, North Carolina, United States
Show Abstract9:00 PM - R7.30
Depinning Transition of Dislocation Line in Oxide Dispersion Strengthened Alloys.
Botond Bako 1 , Daniel Weygand 2 , Maria Samaras 1 , Wolfgang Hoffelner 1 , Michael Zaiser 3
1 , Paul Scherrer Institute, Villigen-PSI Switzerland, 2 , IZBS University of Karlsruhe, Karlsruhe Germany, 3 , University of Edinburgh, Edinburgh United Kingdom
Show AbstractDiscrete dislocation dynamics simulations are used to investigate the dynamics of a driven dislocation line interacting with randomly distributed, incoherent oxide dispersoids that act as pinning centers in a BCC ferritic alloy. The dislocation line undergoes a depinning transition where the order parameter is the mean dislocation line velocity, which increases in the depinning region from zero for external resolved shear stresses beyond a threshold value. The critical stress and critical exponents characterizing the depinning transition are determined numerically, and the observed dynamical behavior is compared with that of a depinning elastic string.
9:00 PM - R7.31
Geometrical and Electronic Structure for Random Grain Boundary of Stainless Steel Systems.
Takahiro Igarashi 1 , Tetsuya Nakazawa 1 , Tomohito Tsuru 1 , Yoshiyuki Kaji 1
1 , Japan Atomic Energy Agency, Ibaraki Japan
Show AbstractSettlement of aging of nuclear reactor is important issue for safe long term operation of light water reactor. Stress corrosion cracking (SCC) is one of the problems of the aging of nuclear reactor, and various types of studies have been carried out with a sight of macroscopic and microscopic view. By experimental study, it was clear that cracks propagate mainly along random grain boundaries. So far, many analyses of stability of grain boundaries have been performed by various computational approaches. However, almost all studies investigated coincident site lattice (CSL), and any random grain boundaries were not investigated. For clarifying the mechanisms of SCC, understanding of random grain boundary properties is significant. In this study, the grain boundary energies of stainless steels systems were analyzed. Especially we aimed at random grain boundaries, and the comparison of the energy between random grain boundaries and CSL was performed. Since random grain boundaries are, however, non-symmetric, there are no explicit definitions of random grain boundaries. In this study, disordered grain boundaries by heat-quenching twisted boundary using molecular dynamics simulation was defined as random grain boundary. The energy analyses of developed random grain boundaries were carried out. Since these systems are constructed by large number of atoms, analyses by first-principle calculation are quite difficult. Then enhanced semi-empirical molecular orbital method was developed, and energy analyses of the grain boundaries were carried out using the developed method in this study. This method enables us to analyze large systems (which have more than 10000 atoms) by partitioning the system to sub-cluster and carrying out molecular orbital analyses of each sub-cluster individually. In the method, all main group elements and transition metals, which include main elements of stainless steel, iron, nickel, and chrome, can be parameterized in the Hamiltonian. Then this approach is appropriate to random grain boundaries of the stainless steel system. The energy analyses of random grain boundaries and CSL by change in chemical composition were performed. It was considered that the difference of the energies between random grain boundaries and CSL related to the strength of atomistic bonding at the grain boundaries.
9:00 PM - R7.32
Molecular Dynamics Study of Thermomigration of Voids in Single Crystal UO2.
Tapan Desai 1 , Paul Millett 1 , Dieter Wolf 1
1 Material Properties and Performance, Idaho National Laboratory, Idaho Falls, Idaho, United States
Show AbstractIt is well known that within a few hours after the startup of a nuclear reactor, the temperature gradient within a fuel element causes migration of voids radially inwards to form a central hole. To understand the atomic processes that control the thermomigration of voids, we performed molecular dynamics simulations on single crystal UO2 with voids (d = 2.2 nm). The system was equilibrated for 1 ns at a temperature (T = 2800 K) well above the oxygen sub-lattice disordering. This high temperature ensured surface diffusion of uranium and oxygen ions. Then, a temperature gradient was applied by supplying additional kinetic energy at the ends of the simulation cell and removing this kinetic energy from the center of the cell. At the end of the simulation run of 19 ns, we found that the voids move towards the hot ends and the void mobility is controlled by the surface diffusion of uranium ions. As the voids migrate, the trailing region on the uranium sub-lattice is completely restored. This work was supported by the DOE-BES Computational Materials Science Network.
9:00 PM - R7.33
Thermal Performance of Deep-Burn Fusion-Fission Hybrid Waste in a Repository.
James Blink 1 , Veraun Chipman 1 , Pihong Zhao 2
1 , LLNL, Las Vegas, Nevada, United States, 2 , LLNL, Livermore, California, United States
Show AbstractConceptual fusion-fission hybrid engines being developed at LLNL can reach extremely high burnup. The resulting nuclear waste is fundamentally different than once-through light water reactor waste currently being considered for disposal in a geologic repository. This paper develops the thermal loads on the repository during the pre-closure and post-closure periods, proposes design and operational features to accommodate the thermal loads, and calculates the resulting temperature histories at key locations.
9:00 PM - R7.34
Radiological Aspects of Deep-Burn Fusion-Fission Hybrid Waste in a Repository.
Henry Shaw 2 , Pihong Zhao 2 , James Blink 1 , Veraun Chipman 1
2 , LLNL, Livermore, California, United States, 1 , LLNL, Las Vegas, Nevada, United States
Show AbstractConceptual fusion-fission hybrid engines being developed at LLNL can reach extremely high burnup. The resulting nuclear waste is fundamentally different than once-through light water reactor waste currently being considered for disposal in a geologic repository. This paper compares the radionuclide composition of the hybrid waste to that from light water reactors, and determines the resulting effects on repository capacity and performance.
9:00 PM - R7.35
Rate Theory Modeling of Irradiation-induced Phosphorus Segregation using First Principles Calculation.
Ken-ichi Ebihara 1 , Masatake Yamaguchi 1 , Hideo Kaburaki 1 , Yutaka Nishiyama 2
1 Center for Computational Science & e-Systems, Japan Atomic Energy Agency, Ibaraki pref. Japan, 2 Nuclear Safety Research Center, Japan Atomic Energy Agency, Ibaraki pref. Japan
Show Abstract Phosphorus segregation under irradiation brings about changes of mechanical properties of nuclear materials. However, the effect of various irradiation parameters, such as temperature, dose rate, and dpa, on the transport process is not clearly understood. It is known that impurity atoms in metals and alloys are transported by migration of vacancies and interstitials which are produced by irradiation; that is the vacancy mechanism and the interstitial mechanism. We take the previous results of phosphorus segregation in nickel alloy [1], and reconsider both mechanisms based on our recent first principles calculations for activation and binding energies of vacancy and interstitial defects. We obtained two remarkable results from our first principles calculation; (1) activation energy of a vacancy, association energy of a vacancy with a phosphorus atom, and phosphorus migration energy through a vacancy are smaller than the reported values [1], (2) one binding configuration between a self interstitial dumbbell and a substitutional phosphorus atom is unstable and not realizable. According to these results, we derived transport coefficients of vacancy, interstitial, and phosphorus atoms from kinetic equations for defect configurations. Then we incorporated the derived transport coefficients to the diffusion-reaction rate equations for vacancy, interstitial, and phosphorous atoms to simulate one-dimensional phosphorus distribution in nickel alloy under ion irradiation. The simulation results show that the dpa value was much smaller than that of the simulation and the experimental result reported in Ref.[1] because phosphorus migration by the vacancy mechanism was much faster due to (1) and phosphorus could not be transported by the interstitial mechanism due to (2). Thus we adjusted vacancy sink strength for suppressing the amount of vacancies and incorporated a term for phosphorous self-migration through interstitial sites to the rate equation instead of the interstitial mechanism. We successfully obtained the phosphorous distribution which is close to the results in Ref. [1] and, furthermore, the present result on the distribution around the peak of vacancy production by nickel ion irradiation is much closer to the experimental result rather than simulation result shown in [1]. The present result clearly indicates that a reduction in vacancies plays a central role on the transport of phosphorus atoms.Reference[1] S. M. Murphy and J.M. Perks, "Analysis of Phosphorus Segregation in Ion-irradiated Nickel", J. Nucl. Mater. 171, pp.360-372 (1990).
9:00 PM - R7.36
A New Magnetic Potential for Ferromagnetic BCC Fe.
Samuele Chiesa 1 , Peter Derlet 1 , Sergei Dudarev 2 , Helena Van Swygenhoven 1
1 ASQ/NUM – Materials Science & Simulation, Paul Scherrer Institut, Villigen PSI Switzerland, 2 Culham Science Centre, EURATOM/UKAEA Fusion Association, Oxfordshire OX14 3DB United Kingdom
Show AbstractThe Magnetic Potential [J. Phys.: Cond. Matt. 17, 7097 (2005), J. Prog. Mater. Sci. 52, 299 (2007)] is an inter-atomic potential for bcc iron that explicitly takes magnetism into account and has been developed to study irradiation damage. The present work reports on the development of a new parameterization of the magnetic potential that is directly fitted to the experimental/ab initio third order elastic constants. Fitting to such material parameters are shown to result in a potential that reproduces well a wide range of anharmonic material properties such as volume expansion and how elastic constants are modified in the presence of defects. Additionally, the magnetic potential has been optimised to produce correctly the ab intio derived properties of a pure <111> screw dislocation through direct control of the associated general planar fault energy surface and the 111 atomic strings displacement energetics. The vacancy and interstitial kink structures that together define the kink pair formation energy – a materials parameter that is known to constitute the rate limiting process in dislocation based bcc plasticity - are investigated and compared to ab initio. We conclude by discussing further strategies for the improvement of the magnetic potential.
9:00 PM - R7.37
Kinetic and Metropolis Monte Carlo Simulations of Ordering in FeCr Alloys.
Evgueny Zhurkin 1 , Perera Roman 2 , Nicolas Castain 3 , Lorenzo Malerba 3 , Marc Hou 4
1 Department of Experimental Nuclear Physics K-89, St.Petersburg State Polytechnical University, St Petersburg Russian Federation, 2 INSTN, CEA Saclay, Gif sur Yvette France, 3 Reactor Materials Research Unit, SCK.CEN, Mol Belgium, 4 Physique des Solides Irradiés et des Nanostructures CP234, Université Libre de Bruxelles, Bruxelles Belgium
Show AbstractMonte Carlo simulations are standard computational methods to study equilibrium and non-equilibrium thermodynamic properties of a system at the atomic level. The Metropolis Monte Carlo (MMC) algorithm is especially suited to drive the system towards thermodynamic equilibrium. The algorithm accounts for all terms that contribute to defining the free energy difference between states: not only chemical, configurational and interfacial, but also due to strain fields. While the configurations visited using MMC do not represent a physical trajectory, kinetic Monte Carlo does, but uses approximations such as assuming rigid lattices. In this work, both methods are used to predict the same ordering properties of FeCr alloys at various compositions and temperatures in the absence of defects. The comparison allows discussing approximations used in the kinetic Monte Carlo approach and a mutual assessment of both methods.
9:00 PM - R7.38
The Structure, Role and Flexibility of Grain Boundaries.
Maria Samaras 1 , Ning Gao 1 , Maximo Victoria 1 2 3 , Wolfgang Hoffelner 1
1 , Paul Scherrer Institiute, Villigen Switzerland, 2 , Lawrence Livermore National Laboratory, Livermore, California, United States, 3 , Polytechnic University of Madrid, Madrid Spain
Show AbstractGrain boundaries (GBs) play an important role in crystalline materials as they control many of the material's physical and chemical properties. At high temperature and during irradiation, GBs act as sinks or sources for interstitials and vacancies. To investigate the role of the GB, it is first necessary to deduce which GB configurations are stable. In this paper we clarify and classify the different types of GBs, their role and flexibility using atomistic simulations. For example, bi-crystal molecular static simulations of iron GBs show that the structure of some GBs is potential dependent. Indeed such potential dependence also encompasses molecular dynamics irradiation simulation of nanocrystalline iron. These issues as well as GB flexibility in an irradiation environment will be presented in this paper.
9:00 PM - R7.39
Mechanism for Kinetically-Driven Point-Defect Clustering in MgO by Molecular-Dynamics Simulation.
Dilpuneet Aidhy 1 , Paul Millett 2 , Simon Phillpot 1 , Dieter Wolf 2
1 Materials Science and Engineering, University of Florida, Gainesville, Florida, United States, 2 Material Sciences Department, Idaho National Laboratory, Idaho Falls, Idaho, United States
Show AbstractRadiation damage involves first Frenkel-pair (FP) formation during the initial ballistic phase, and then point-defect recombination and clustering during the kinetic phase in addition to electronic and microstructural changes. The ballistic phase occurs on the picosecond time scale and much attention has been paid to capturing the collision cascade using molecular-dynamics (MD) simulation. The recombination phase, however, takes much longer time and the real lasting damage to the material occurs during this time. Therefore, it is extremely important to understand these kinetically-driven radiation-damage mechanisms. Unfortunately, due to time limitation, MD until now has not been able to capture this phase completely.Here, we have chosen MgO as a model material in which to demonstrate the power of a novel approach for understanding the kinetic phenomena associated with the recombination and clustering processes. Our approach involves random creation of a high density of FPs and then allowing recombination and cluster formation according to the equilibrium kinetics of the system. From previous MD simulations of collision cascades, it has been shown that the clustering of the MgO interstitials during the kinetic phase is only weakly dependent on the initial configuration of the defects created during the ballistic phase[1]. Our approach is able to replicate the nature and size of the clusters (di-, tri- and larger interstitial clusters) that were previously observed[1]. In addition, interstitial clusters containing both anions and cations are found to be formed, structures that have not previously been identified. Further, we find that to stabilize these interstitial clusters, the lattice atoms participate in cluster formation by creating additional Frenkel pairs. As more and more interstitials precipitate into the clusters, a new interstitial rocksalt sub-lattice is formed inside the host lattice. This work was supported by DOE NERI contracts DE-FC07-07ID14833 and DE-FC07-05ID14649 and by the DOE-BES Computational Materials Science Network (CMSN). 1B.P. Uberuaga, R.Smith, A.R. Cleave, F. Montalenti, G. Henkelman, R.W. Grimes, A.F. Voter, K.E. Sickafus, Structure and Mobility of Defects Formed from Collision Cascades in MgO, Physical Review Letters, Vol. 91, 11, 2004.
9:00 PM - R7.4
Hydrogen Permeation in Tungsten and Copper Langmuir Probes at ISTTOK.
Daniela Nunes 1 , Rodrigo Mateus 1 , Patrícia Carvalho 1 2 , Augusto Deus 2 , José Correia 3 , Horácio Fernandes 1 , Carlos Silva 1 , Nuno Franco 4 , Eduardo Alves 4 , Rui Gomes 1
1 Euratom/IST, Instituto de Plasmas e Fusão Nuclear, Instituto Superior Técnico, Lisbon Portugal, 2 Euratom/IST, Departamento de Engenharia de Materiais, Instituto Superior Técnico, Lisbon Portugal, 3 Departamento de Materiais e Tecnologias de Produção, LNEG, Lisbon Portugal, 4 Instituto Tecnológico e Nuclear, ITN, Sacavém Portugal
Show AbstractThe high melting point, high threshold for sputtering and low tritium inventory of tungsten render this material potentially suitable for high flux components and high-power density structural applications in fusion reactors. Copper alloys are considered appropriate heat-sink materials for first wall panels due to their favorable thermal conductivity, mechanical strength and radiation resistance.Energetic hydrogen ions are able to penetrate the surface barrier of metals and this ion driven permeation can exceed gas driven permeation by several orders of magnitude. The solubility of hydrogen isotopes in tungsten and copper is low and hydrogen atoms are likely to come out of solution at traps such as vacancies, voids, and grain boundaries, where they achieve a lower potential energy than that attained at crystalline sites. The traps eventually become saturated and, even for the low neutral gas pressure in the fusion chamber, untrapped hydrogen atoms tend to diffuse deeper into the material. At the saturated traps, molecular recombination tends to occur resulting in bubbles and blisters with increasing internal pressure. At enhanced diffusivity conditions the system can reduce the interfacial free energy by bubble and blister coarsening.Commercially pure W and Cu wires have been exposed at the ISTTOK tokamak edge plasma to power densities of ~0.1 - 1 MWm-2 for ~200 discharges of 30 ms where Te~Ti = 10-40 eV and n = 0.5-2x1018 m-3 with pulse intervals of 10 min in 10-4 torr H2 atmospheres. The microstructure of both materials revealed recrystallization and strong grain growth at the most severely exposed regions. A high density of small intergranular bubbles was found at the intermediate regions, whereas a low number of large intergranular bubbles present at the most severely exposed regions, as a result of a coarsening process. The internal surface of large bubbles displayed slip bands induced by the high H2 pressure. Hydrogen saturation was assessed through the lattice parameter and Young modulus variations across longitudinal cross-sections of the probes.The results indicate that bubble formation in tungsten and copper first wall components can be expected to occur and strategies for minimization of the phenomenon need to be implemented.
9:00 PM - R7.40
Molecular Dynamics Simulation of Point-Defect Kinetics and Annihilation in Nanocrystalline Metals.
Paul Millett 1 , Dilpuneet Aidhy 2 , Vesselin Yamakov 3 , Simon Phillpot 2 , Dieter Wolf 1
1 , Idaho National Laboratory, Idaho Fall, ID, Idaho, United States, 2 Department of Materials Science and Engineering, University of Florida, Gainsville, Florida, United States, 3 , National Institute of Aerospace, Hampton, Virginia, United States
Show AbstractDue to their very high density of nucleation centers coupled with the short diffusive lengths involved, nanostructured materials should be able to respond particularly quickly to any transient imbalances in their inherent thermal-equilibrium point-defect concentrations due to, for example, collision cascades during irradiation or temperature fluctuations. Here, we examine the kinetics of super-saturated point-defect (i.e., vacancy and self-interstitial) concentrations within nano-sized grains of bcc-molybdenum using molecular-dynamics simulations carried out at high temperatures (in order to activate point-defect migration). These simulations provide atomic-level insights into the nature of GB source/sink behavior, Frenkel pair recombination rates within the grain interiors, as well as the overall time scales related to re-equilibration of defect concentrations following a uniformly-imposed supersaturation. Furthermore, these simulations provide a basis to clarify the results of recent experiments focused on understanding irradiation tolerance in nanostructured materials. To this end, we examine the efficiency of GB sinks during extended periods of absorption of high fluxes of both intrinsic point-defect species.
9:00 PM - R7.41
Multiple-Beam Irradiation Effects in F82H Electron Beam Welded Joint.
Naoki Shiomi 1 , Naoyuki Hashimoto 1 , Somei Ohnuki 1 , Hiroshi Kinoshita 1
1 Material Science, Hokkaido University, Sapporo Japan
Show Abstract9:00 PM - R7.42
Formation Mechanism of Solute Clusters in Low Copper or Copper Free Ferritic Model Alloys and in a RPV Steel under Neutron Irradiation: A Combined Theoretical and Experimental Study.
Estelle Meslin 1 , Bertrand Radiguet 2 , Philippe Pareige 2 , Alain Barbu 1 , Hideo Sakasegawa 1
1 SRMP, CEA Saclay, Gif sur Yvette France, 2 , University of Rouen, Saint Etienne du Rouvray France
Show Abstract9:00 PM - R7.43
Interface Chemistry and Structure of Inconel 690 in Corrosion.
Joseph Fricano 1 , Joseph Marra 1 , Bilge Yildiz 1
1 Nuclear Science and Engineering, MIT, Cambridge, Massachusetts, United States
Show Abstract9:00 PM - R7.5
Characterization of Oxide and Carbide Particles in a ODS Fe12%Cr Alloy.
Emmanuelle Marquis 1 , Sergio Lozano-Perez 1 , Vanessa de Castro 1
1 Department of Materials, University of Oxford, Oxford United Kingdom
Show AbstractOxide dispersion strengthened (ODS) reduced activation ferritic-martensitic steels (RAFMS) are promising structural materials for the first wall and blanket of future fusion reactors. The analysis of the microstructures after annealing and irradiation is essential to confirm the applicability of these materials. Current steels, such as the ODS EUROFER 97 alloy, exhibit complex microstructures due to both the production route and the presence of multiple alloying elements. Therefore, in the present work, a model ODS system based on a Fe-12wt.%Cr binary alloy was chosen in order to isolate the role of the Y2O3 particles and achieve a better understanding of the microstructural evolution in damage development. The ODS Fe-Cr alloy was tempered at 750oC for 4 to 96 hours. Characterization using transmission electron microscopy and atom probe tomography focused on the second phase particles and their evolution during tempering. An inhomogeneous distribution of dispersion of yttrium-rich particles with sizes between 2 and 50nm was observed. A Cr-rich shell was observed around some particles and carbon was sometimes found at their interfaces as well. Larger Cr2O3 oxides and MxOy (M=Cr, Y) particles with varying composition were found and M23C6 (M=Cr, Fe) carbides were observed at grain boundaries and sometimes attached to oxide particles. Detailed structural and compositional analyses will be presented. The authors acknowledge T. Leguey, A. Muñoz, M. A. Monge and R. Pareja for providing the alloys. This research was supported by FP6 Euratom Research and Training Programme on Nuclear Energy and by the Engineering and Physical Sciences Research Council (EPSRC) under grant number EP/077664.
9:00 PM - R7.6
Effects of Multi-Beam Irradiation on Defect Formation in Fe-Cr Alloys.
Hitoshi Seto 1 , N. Hashimoto 1 , S. Ohnuki 1 , H. Kinoshita 1
1 division of materials science and engineering, graduate school of engineering, Hokkaido University, Sapporo Japan
Show AbstractReduced-activation ferritic/martensitic steels, which have high irradiation resistance, are candidates for the material of first wall and/or blanket of International Thermonuclear Experimental Reactor (ITER). Hydrogen and/or helium created by nuclear transmutation reaction could change the migration energy of point defects such as vacancy and interstitial, which influence evolution of irradiation-induced dislocation loops. This study aims to investigate the effect of helium on migration energy of vacancy or interstitial by in-situ observation of an electron-irradiated Fe-8Cr model alloy, which is fundamental composition of the reduced-activation ferritic/martensitic steel developed by Japan Atomic Energy Agency (JAEA).TEM specimens of Fe-8Cr were irradiated under two modes: simple electron irradiation and dual-beam-irradiation by electron and helium ion, and in-situ observed in the high voltage electron microscope operated at 1250 kV at 300-500 °C. The accelerating voltage of helium ion was 100 kV and the implanting concentration was about 5 appm/dpa. At all temperature, the diameter and number density of dislocation loops were increased as a function of dose on both specimens. It appears that the dual-irradiated specimens showed greater temperature dependence on the growth rate of dislocation loop than that of electron-irradiated. On the other hand, there was little difference on temperature dependence of the saturated loop number density. It is known from past studies that the migration energy of vacancy can be estimated from growth rate of dislocation loop. From this in-situ experiment, it is indicated that net migration energy of vacancies was raised up due to trapping by helium, resulted in greater temperature dependence of the growth rates.
9:00 PM - R7.7
Interaction between Dislocation Sliding and Damage Structure in Ion-irradiated Stainless Steels.
Terumitsu Miura 1 , Katsuhiko Fujii 1 , Koji Fukuya 1
1 , Institute of Nuclear Safety System, Incorporated, Fukui Japan
Show AbstractThe interaction between dislocation sliding and damage structures in irradiated stainless steels was investigated for Fe ion-irradiated type 316 and 304 stainless steels (SS). The difference on the interaction between Fe and He ion-irradiations was also examined.Small tensile test specimens with the gage section of 9×3×1mm were prepared from solution annealed 316SS and 304SS. After mechanical and electro-polishing, one side of the gage section was irradiated with 2.8MeV Fe2+ ions at 300°C up to 10dpa. The damage rate was about 5×10-4/s at damage peak depth. The microstructures of damage regions were observed by transmission electron microscope (TEM) and hardness in the damage regions were measured using an ultra-micro hardness tester. The specimens were tensiled at a strain rate of 10-7/s to 2% plastic strain at 300°C in argon gas. After the tensile tests, slip line spacing for coarse slip steps on the surface was measured using an optical microscope and near-surface deformation microstructures were observed by TEM.Dislocation loops and voids were observed as damage structures. The loop size increased and loop density was almost constant whereas the void size was constant and void density increased and then saturated as the irradiation dose increased. The average diameters of dislocation loops and voids at damage peak depth in 316SS irradiated to 10dpa were about 14nm and 1nm, respectively; whose number densities were about 1×1023/m3 and 7×1023/m3, respectively. The sizes and number densities of the irradiation defects in 304SS were close to 316SS. The hardness of damage region increased as the irradiation dose increased. The hardness of 316SS and 304SS were close before and after irradiations.The slip line spacing in 316SS and 304SS were close before irradiations. However, the spacing in 316SS increased by about 10% at 10dpa whereas the spacing in 304SS increased by about 30% at 5dpa. Dislocation channels were observed in the damage region, which were formed due to sweeping the irradiation defects by dislocation sliding from the matrix. Some of the channels were blocked in the damage region whereas others penetrated through the damage region and formed coarse surface slip steps. Thus, the slip line spacing on the surface widened in a consequence of preventions of dislocation sliding by the damage structures.The hardness increase and slip line spacing after the Fe and He irradiations could be compared by using the shear strength increase calculated from microstructural data using the Orowan model. The hardness increase and slip line spacing were small for Fe ion-irradiated specimens because the number densities of irradiation defects were low compared in He ion-irradiated specimens.The prevention of dislocation sliding by damage structures depended on kind, diameter and number density of the irradiation defects. The prevention became strong in 304SS compared in 316SS, probably due to lower stacking fault energy in 304SS.
9:00 PM - R7.8
W-nDiamond/Cu-nDiamond Nanostructured Composites for Fusion Devices.
Daniela Nunes 2 , Vanessa Livramento 1 , José Correia 1 , Nobumitsu Shohoji 1 , Rodrigo Mateus 2 , Patrícia Carvalho 2 3 , Horácio Fernandes 2 , Carlos Silva 2 , Eduardo Alves 4 , K. Hanada 5 , E. Osawa 6
2 Euratom/IST, Instituto de Plasmas e Fusão Nuclear, Instituto Superior Técnico, Lisbon Portugal, 1 Departamento de Materiais e Tecnologias de Produção, LNEG, Lisbon Portugal, 3 Euratom/IST, Departamento de Engenharia de Materiais, Instituto Superior Técnico, Lisbon Portugal, 4 Instituto Tecnológico e Nuclear, ITN, Sacavém Portugal, 5 National Institute of Advanced Industrial Science and Technology , AIST, Ibaraki Japan, 6 NanoCarbon Research Institute, Shinshu University, Nagano Japan
Show AbstractA novel material design for nuclear fusion reactors is proposed based on W-nDiamond nanostructured composites. Generally, a microstructure refined to the nanometer scale improves the mechanical strength due to modification of plasticity mechanisms. Moreover, high specific grain-boundary area raises the number of sites for annihilation of radiation induced defects. However, the low thermal stability of fine-grained and nanostructured materials demands the presence of particles at the grain boundaries that can retard coarsening by a pinning effect. As a compromising solution, a composite approach can be promising in the field of nanostructured materials. The hardness of diamond renders nanodiamond dispersions excellent reinforcing and stabilization candidates and, in addition, diamond has extremely high thermal conductivity. Consequently, W-nDiamond nanocomposites are promising candidates for thermally stable first-wall materials. The proposed design involves the production of W/W-nDiamond/Cu-nDiamond/Cu layered castellations. The W, W-nDiamond, Cu-nDiamond and Cu layers are produced by mechanical alloying followed by a consolidation route and interlayer welding that combines hot rolling with spark plasma sintering. Modulation of the processing parameters allows controlling the extent of conversion of diamond into tungsten carbide, which was tuned to yield virtually nil carbide content. Microstructural features and mechanical properties of the as-produced material and after exposure to the ISTTOK edge plasma are presented.
9:00 PM - R7.9
Effect of Irradiation on Grain Boundary Chemistry in Oxide-Dispersion Strengthened Fe-Cr Alloys.
Emmanuelle Marquis 1 , Ceri Williams 1 , Alfred Cerezo 1
1 Department of Materials, University of Oxford, Oxford United Kingdom
Show AbstractUnderstanding the behaviour of oxide-dispersion strengthened (ODS) ferritic martensitic steels under irradiation is of prime importance in the design of future fusion reactors. It has been observed, in austenitic steels, that grain boundary chemistry can be significantly affected by irradiation with chromium depletion occurring at grain boundaries that would otherwise exhibit segregation under thermodynamic equilibrium. A possible mechanism relates to the substitutional large atoms diffusing away by vacancy exchange (also called the inverse Kirkendal mechanism) and smaller interstitial atoms being accommodated by vacancies in dumbbell configurations and segregating at the defect sinks.Here, the effect of heavy-ion implantation on grain boundary chemistry was investigated using atom-probe tomography in a model ODS Fe-12 wt.% Cr alloy and a ODS Euroder 97 steel. Chromium and carbon segregation at grain boundaries is found in the annealed alloys. After irradiation at 300oC to less than 5dpa, a W-shaped profile develops at some boundaries. After irradiation at 500oC, complete Cr depletion of some grain boundaries is observed, while Cr segregation is maintained at other boundaries. The three-dimensional reconstructions reveal a complex distribution of Cr segregation and depletion at grain boundaries of varying character. The effect of temperature, chemistry, and dose will be discussed.The authors acknowledge V. de Castro, T. Leguey, A. Muñoz, M. A. Monge and R. Pareja for providing the ODS Fe12Cr alloy and R. Lindau (FZK, Germany) for the Eurofer 97 steel. This research is supported by the Physical Sciences Research Council (EPSRC) under grant number EP/077664.
Symposium Organizers
Chu Chun Fu CEA-Saclay
Akihiko Kimura Kyoto University
Maria Samaras Paul Scherrer Institute
Magdalena Serrano de Caro Lawrence Livermore National Laboratory
Roger E. Stoller Oak Ridge National Laboratory
R8: Defect Evolution
Session Chairs
Simon Phillpot
Julie Tucker
Thursday AM, December 04, 2008
Independence W (Sheraton)
9:15 AM - **R8.1
Nucleation and Growth of He-vacancy Complexes in α-Fe: Effect of Carbon and Comparison to Experiments.
Christophe Ortiz 1 2 , M. Caturla 1 , C. Fu 3 , F. Willaime 3
1 , Universidad de Alicante, Alicante Spain, 2 , Laboratorio Nacional de Fusion por Confinamento Magnetico CIEMAT, Madrid Spain, 3 , Service de Recherches de Metallurgie Physique CEA/Saclay, Gif-sur-Yvette Cedex France
Show AbstractUnderstanding He migration and its interaction with defects produced by irradiation such as vacancies or self-interstitials, as well as the microstructure, dislocations and grain boundaries, is particularly important for fusion applications. The levels of He produced in fusion reactors will be significantly higher than those achieved under fission conditions, and therefore, new effects could be expected. Materials such as ferritic/martensitic steels have been selected for this application due to their low swelling rates . However they suffer from embrittlement and the role of He on this phenomenon is not yet known. In order to elucidate the large number of processes that take part during the irradiation of a material it is necessary to combine model development, based on a multiscale approach, and tailored experiments to validate these models.During the past years we have been working on developing a model for He migration and evolution in α-Fe in the presence of defects produced by irradiation. This model consists of a combination of calculations using first principles density functional theory (DFT) to obtain basic parameters, such as migration and binding energies of defects, and rate theory (RT) and/or object kinetic Monte Carlo (OKMC) models to reproduce the long term evolution of these defects. Using a RT model based on DFT calculations we have studied the desorption of He from α-Fe under different implantation and annealing conditions. These results have been compared with experimental measurements and showed the importance of taking into account, at least in a material such as α-Fe, the presence of impurities, in particular carbon. Supported by DFT data, simulations evidenced that carbon not only interacts with vacancies, which retards their migration in α-Fe, but also affects He desorption. The results of these simulations as well as their consequences will be presented in detail. We are currently in the process of performing new experimental measurements of He implantation and desorption to validate the model, in particular the effects of carbon. Also the model is being extended to include the interaction of He with grain boundaries and formation of bubbles along these defects. At the same time, we will also discuss the limitations of the RT model, which is based on a mean-field approximation. Comparisons with OKMC calculations with the same set of input parameters allow us to check the range of validity of RT models under different conditions.
9:45 AM - R8.2
Modeling Microstructural Evolution of Ferritic Martensitic Steels under High Energy Ion Irradiation.
Francois Gallet 1 , Brian Wirth 1 , Donghua Xu 1
1 Nuclear Engineering, UC Berkeley, Berkeley, California, United States
Show AbstractFuture fission and fusion reactors will require advanced materials to meet design goals of higher temperatures, higher radiation doses and higher energy neutron spectrum. Indeed, core internals, fuel cladding (typically austenitic, ferritic or martensitic alloys) and first wall materials will experience performance degradation involving hardening and embrittlement, swelling, creep and must resist helium embrittlement. The performance changes are intimately related to microstructural changes, such as the formation of point defect clusters and dislocation loops, as well as radiation enhanced diffusion resulting in solute segregation and precipitation. This presentation will describe the development of a temporally and spatially dependent rate theory, or cluster dynamics, model that describes the evolution of self-interstitial (mainly interstitial loops) and vacancy clusters (voids) in ferritic-martensitic steels under irradiation. The spatial dependence of the model allows variation in sink density, as occurs near precipitates, grain boundaries and dislocations, and provides flexibility in the number of mobile point defect clusters. The initial simulations have focused on high-energy ion irradiation of ferritic-martensitic steels to compare with recent in-situ transmission electron microscopy irradiation studies. The results presented will also investigate the proposed formation mechanicsms of a <100> dislocation loop in irradiated ferritic-martensitic alloys.
10:00 AM - R8.3
Monte Carlo Simulations of Irradiated Materials on the Reactor Timescale.
Vasily Bulatov 1 , Aleksandar Donev 1 , George Gilmer 1
1 , Lawrence Livermore National Laboratory, Livermore, California, United States
Show AbstractMaterials irradiated by high-energy particles constitute an important example of systems far from thermodynamic equilibrium. Copious quantities of atomic defects produced by irradiation give rise to complex diffusion-controlled processes defining the evolution of material microstructure. The time scales on which this evolution takes place range from nanosecond time intervals of fast atomic diffusion to tens of years of material work life in nuclear reactors. Here we present a new method for kinetic Monte Carlo simulations that encompasses all relevant time scales and connects the fundamental atomistic mechanisms directly to the long-term damage accumulation. The new simulation method is tested on a simple model of α-iron for a range of temperatures and irradiation dose rates. The results suggest an approximate scaling relationship by which material damage observed in accelerated irradiation tests can be extrapolated to predict how the same material would resist irradiation over much longer time scales of nuclear reactors.
10:15 AM - R8.4
A Diffuse-interface Phase-field Model for Simulation of Void Growth in Irradiated Materials.
Srujan Rokkam 1 , Anter El-Azab 1 2 , Paul Millet 3 , Dieter Wolf 3
1 Mechanical Engineering Department, Florida State University, Tallahassee, Florida, United States, 2 School of Computational Science, Florida State University, Tallahassee, Florida, United States, 3 Materials Science Department, Idaho National Laboratory, Idaho Falls, Idaho, United States
Show AbstractStructural materials in nuclear reactor systems are subject to excessive point defect generation by virtue of atomic collision cascades. The diffusion and interaction of these point defects leads to profound changes in the microstructure of the material. A very important microstructural aspect in irradiated materials is the nucleation and growth of voids, which can cause volumetric swelling and detrimental dimensional changes. The conventional approach for modeling the void swell problem considers the void nucleation and growth stages separately, which are treated using classical nucleation theory and rate theory, respectively. We present a diffuse-interface phase-field model for nucleation and growth of voids in irradiated materials, in which the processes of growth and nucleation are treated simultaneously in a spatially resolved fashion. The defect fluxes and the defect density modulations are formulated using Cahn-Hilliard type description for the vacancy and interstitial concentration fields. The void growth dynamics is obtained in terms of the evolution of a non-conserved order parameter field, whose evolution is prescribed by a phenomenological Allen-Cahn type equation. The model also accounts for the effect of applied stress, cascade-induced and thermally-induced fluctuations, vacancy-interstitial recombination, and interaction of vacancies and interstitials with lattice sinks. Illustrative results of model capabilities are presented using the case of a pure metal as an example.
10:30 AM - **R8.5
Accelerated Kinetic Monte Carlo Simulations of Defect Accumulation in Irradiated Systems.
Jaime Marian 1 , Enrique Martinez 2 , J. Manuel Perlado 3
1 , Lawrence Livermore National Laboratory, Livermore, California, United States, 2 , CEA, Saclay France, 3 , Universidad Politecnica de Madrid, Madrid Spain
Show AbstractSerial kinetic Monte Carlo (kMC) calculations are systematically approaching their saturation limit in terms of computational capacity. In the context of radiation damage, standard serial kMC methods have provided defect density time evolutions of several tenths of a dpa. This is clearly insufficient to model damage accumulation in new-generation nuclear systems, where doses and temperatures are much higher than those found in current LWRs. Here we present calculations of defect accumulation in metals using a novel parallel kMC (pkMC) algorithm in conditions representative of those expected in fusion materials. Our pkMC advances time synchronously, ideally K times as fast a serial kMC (where K is the number of processors). This ideal scaling can be can be reached with appropriate spatial decompositions based on global rules of total rate equipartition. The parallel efficiency and the error incurred are also discussed.
R9: Fuels
Session Chairs
Jeremy Bischoff
Emmanuelle Marquis
Thursday PM, December 04, 2008
Independence W (Sheraton)
11:15 AM - **R9.1
Discovery and Design of Advanced Nuclear Fuels.
Marius Stan 0
0 , Los Alamos National Laboratory, Los Alamos, New Mexico, United States
Show Abstract11:45 AM - R9.2
Radiation Damage Effects in Uranium-Bearing Delta-phase Oxides.
Ming Tang 1 , James Valdez 1 , Blas Uberuaga 1 , Yongqiang Wang 1 , Kurt Sickafus 1 , Kiel Holliday 2 , Ken Czerwinski 2
1 Materials Science & Technology, Los Alamos National Laboratory, Los Alamos, New Mexico, United States, 2 Radiochemistry Group, University of Nevada, Las Vegas, Munich Germany
Show AbstractActinide oxides have been used as nuclear fuel since the first controlled chain reaction in 1942. UO2 and mixed oxides (MOX) exhibit exceptional resistance to irradiation damage and are therefore still the predominant fuel for thermal power reactors. To develop advanced nuclear fuel or waste forms, the search for radiation tolerant materials has been an area of intense research in recent years.The purpose of this study is to examine the radiation damage behavior of complex actinide oxides whose structures are derivatives of the fluorite structure which is the structure of UO2. The particular actinide compounds of interest in this report are the series of uranium-bearing delta-phase oxides of A6U1O12 (A = Gd to Lu, and Y). These compounds have a crystal structure known as the delta (δ) phase, a structure with rhombohedral symmetry belonging to space group R-3 , which is closely related to the cubic fluorite structure. We have recently shown that certain zirconate δ-phase oxides (4:3:12 stoichiometry) are highly radiation tolerant; in particular, they are especially resistant to amorphization transformations. 6:1:12 compounds are the most interesting delta-phase of all because it is the only known delta composition in which both the cation and anion sublattices are fully-ordered. Polycrystalline Y6U1O12, Gd6U1O12, Ho6U1O12, Yb6U1O12, and Lu6U1O12 samples were synthesized from mixtures of uranium oxide (UO2) and sesquioxide compounds (such as Y2O3 and the rare-earth sesquioxides from Gd2O3 to Lu2O3), then irradiated with 300 keV Kr++ ions to fluences up to 2x1020 ions/m2 at cryogenic temperature (100 K). Ion irradiation damage effects in these samples were examined using grazing incidence X-ray diffraction (GIXRD) and cross-sectional transmission electron microscopy (TEM). Based on our previous studies involving certain zirconate δ-phase compounds, we anticipated that uranate compounds will be resistant to radiation-induced amorphization and undergo an ordered rhombohedral to disordered cubic fluorite phase transformation. Our preliminary results indicate that the compound Y6U1O12 is resistant to amorphization to a displacement damage dose of 50 displacements per atom (dpa), and additionally Y6U1O12 appears to resist cation and anion disordering. Factors influencing the resistance of to irradiation-induced amorphization in these uranium-bearing delta-phase oxides will be discussed in terms of cation radius ratio, unit cell volume, and cation antisite formation energy (an indictor of tendency to undergo an order to disorder transformation). Finally, we will discuss results of density functional theory (DFT) calculations based on the Vienna Ab Initio Simulation Package (VASP) using the projector augmented wave (PAW) method, wherein we calculated the cation antisite formation energies for these compounds.
12:00 PM - R9.3
Radiation Effects in HfO2/MgO Dispersion Nuclear Fuel Surrogate.
Igor Usov 1 , James Valdez 1 , Robert Dickerson 1 , Marylin Hawley 1 , David Dvelin 1 , Yongqiang Wang 1 , Kurt Sickafus 1
1 , Los Alamos National Laboratory, Los Alamos, New Mexico, United States
Show AbstractThe objective of our study is to investigate physical and chemical processes in dispersion nuclear fuels subjected to wide range of radiation damage fluences. The dispersion nuclear fuel consists of fuel particles immersed in a non burnable host matrix. Tailoring the microstructure and geometry of such a dispersion fuel may improve its radiation stability and radically ease the separation of fission products. For the sake of convenience, a surrogate for the dispersion fuel was fabricated by electron beam evaporation in thin film form, HfO2/MgO/HfO2. The MgO layer is the host matrix that can be readily separated from the surrogate fissile particles represented by HfO2. This trilayer composite was irradiated with 10 MeV Au3+ ions over a wide fluence range from 1x1013 to 1x1016 cm-2 at room temperature. Microstructural changes were examined by using transmission electron microscopy (TEM) and grazing incidence X-ray diffraction (GIXRD). Also, the radiation-induced redistribution of composite elements was analyzed by Rutherford backscattering spectroscopy (RBS). Finally, the effect of radiation damage on MgO dissolution will also be discussed.
12:15 PM - R9.4
An Assessment of Layer Development at the Fuel/Cladding Interface During Irradiation of Metallic SFR Fuel Elements.
Dennis Keiser 1 , James Cole 1
1 , Idaho National Laboratory, Idaho Falls, Idaho, United States
Show AbstractTo investigate fuel cladding chemical interaction in irradiated metallic nuclear fuels, diffusion couple experiments have been performed using prototypic metallic fuels with additions of noble metal and lanthanide fission product components and stainless steel claddings. The developed interaction zones have been characterized using SEM/EDS/WDS to determine the development of phases and the interdiffusion behavior of specific fuel, cladding, and fission product components. The formed diffusion structures have been compared to actual diffusion zones that form in irradiated metallic SFR fuels. This talk will discuss how the structures compare between the diffusion couple-generated diffusion zones and the irradiated metallic nuclear fuels. It was found that similarities exist between the phase development and interdiffusion behavior in the annealed diffusion couples and the irradiated fuels. Comments can be made about the types of phases that will form when different types of fuel and claddings interact and where specific fuel, cladding, and fission product components will be found near the fuel/cladding interface.
12:30 PM - R9.5
Sintering Kinetics of Ln Doped dUO2.
S. Yates 1 2 , C. Stanek 1 , K. McClellan 1 , D. Byler 1 , J. Nino 2 , H. Volz 1 , R. Houlton 1 , K. Desai 3 , R. Grimes 3
1 Materials Science and Technology Division, Los Alamos National Laboratory, Los Alamos, New Mexico, United States, 2 Department of Materials Science and Engineering, University of Florida, Gainesville, Florida, United States, 3 Department of Materials, Imperial College, London United Kingdom
Show AbstractIf spent nuclear fuel is reprocessed electrochemically, it is anticipated that a non-negligible fraction of lanthanides (Ln) will be present in transmutation fuel precursors. The presence of these “tramp” lanthanides may affect both oxide fuel synthesis and performance. In order to eventually qualify and license a lanthanide bearing transmutation fuel, the effects on fuel synthesis and performance must be understood. In this study, we investigated the effect of lanthanide additions on the sintering kinetics of depleted UO2. Specifically, we performed high temperature (1600 °C) dilatometry experiments for a range of compositions, ranging from 0.75 – 8 wt% Ln. The lanthanide composition considered (consisting of Ce, La, Nd and Pr) was determined by estimating the fraction of lanthanides remaining after typical electrochemical reprocessing. Furthermore, we used atomic scale simulation to aid in the interpretation of the dilatometry results. In particular, we compared the predicted lattice parameter variation of Ln doped UO2 as a function of incorporation mechanism (i.e. compensated by either oxygen vacancies or U5+), to x-ray diffraction results for the sintered samples. We will discuss the integration of the experimental and modeling results to gain insight into the relationship between Ln dopant incorporation mechanism, oxygen stoichiometry, and the sintering kinetics.
12:45 PM - R9.6
Experimental Determination of Metal Fuel Point Defect Parameters.
Mike Fluss 1 , Scott Mccall 1
1 , Lawrence Livermore National Laboratory, Livermore, California, United States
Show AbstractNuclear metallic fuels are one of many options for advanced nuclear fuel cycles because they provide dimensional stability, mechanical integrity, thermal efficiency, and irradiation resistance while the associated pyro-processing is technically relevant to concerns about proliferation and diversion of special nuclear materials. In this presentation we will discuss recent success that we have had in studying isochronal annealing of damage cascades in Pu and Pu(Ga) arising from the self-decay of Pu as well as the annealing characteristics of non-interacting point defect populations produced by ion accelerator irradiation. Comparisons of the annealing properties of these two populations of defects arising from very different source terms are enlightening and point to complex defect and mass transport properties in the plutonium specimens which we are only now starting to understand as a result of many follow-on studies. More importantly however, the success of these measurements points the way to obtaining important mass transport parameters for comparison with theoretical predictions or to use directly in existing and future materials modeling of radiation effects in nuclear metallic fuels. The way forward on such measurements and the requisite theory and modeling will be discussed.This work performed under the auspices of the U.S. Department of Energy by Lawrence Livermore National Laboratory under Contract DE-AC52-07NA27344.
R10: Fuels Modeling
Session Chairs
Thursday PM, December 04, 2008
Independence W (Sheraton)
3:00 PM - **R10.1
Microscopic Processes in Oxide Fuels: Insights from Simulation.
Simon Phillpot 1 , Taku Watanabe 1 , Dilpuneet Aidhy 1 , Pankaj Nerikar 1 , James Tulenko 2
1 Department of Materials Science and Engineering, University of Florida, Gainesville, Florida, United States, 2 Department of Nuclear and Radiological Engineering, University of Florida, Gainesville, Florida, United States
Show AbstractWe use atomic-level and electronic-level simulation methods to characterize various properties of UO2 fuel. In particular, the ability of atomic-level simulations to determine the thermal thermal-transport properties, including the effects of off-stoichiometry and microstructure is assessed. We also illustrate how such microscopic simulations can provide input to a fuel-performance code, and assess the fidelity of the resulting predictions of materials performance. Electronic-structure calculations are used to characterize the energetics of intrinsic and extrinsic defect formation and migration in UO2. Preliminary results from simulations of radiation damage in polycrystalline UO2 are also discussed. This work was funded by DOE-NERI Awards DE-FC07-05ID14647, DE-FC07-05ID14649 and DE-FC07-07ID14833, and by the DOE-BES Computational Materials Science Network (CMSN).
3:30 PM - R10.2
Ab Initio Modeling of Fission Products Incorporation and Segregation in Alkaline Earth Metal Oxides, Hafnium Dioxide and Their Interfaces.
Xiang-Yang Liu 1 , Blas Uberuaga 1 , Christopher Stanek 1 , Kurt Sickafus 1
1 Materials Science and Technology Division, Los Alamos National Laboratory, Los Alamos, New Mexico, United States
Show AbstractIn order to close the nuclear fuel cycle, advanced concepts for separating out fission products are necessary. One method for doing this is a dispersion fuel form in which a fissile core is surrounded by an inert matrix that captures and immobilizes the fission products from the core. If this inert matrix can be easily separated from the fuel, via e.g. solution chemistry, the fission products can be separated from the fissile core. We examine a surrogate dispersion fuel composition, in which HfO2 is a surrogate for the fissile core and alkaline earth metal oxides are used as the inert matrix. The questions of fission products incorporation in these oxides and possible segregation behavior at interfaces are considered here. Density functional theory based calculations of fission product elements (Xe, Sr, and Cs) in these oxides and hafnium dioxide are carried out. We find smaller incorporation energy in hafnia than in MgO for all three elements. We also find such trend in incorporation energy is reversed or reduced for alkaline earth metal oxides with large cation sizes. Model interfacial calculations show strong segregation tendency from MgO to MgO/HfO2 interfaces. Further calculations using empirical potentials are carried out to calibrate the modeling work.
3:45 PM - R10.3
Ab Initio Studies of Point Defects and Stability of Fission Products in Uranium Dioxide.
Alain Pasturel 1
1 , CNRS, Grenoble France
Show AbstractThe knowledge of the behaviour of fission products (FPs) in the nuclear fuel, UO2, is very important for nuclear safety considerations also for understanding the evolution of the fuel properties under irradiation. Due to the difficulty of performing and interpreting experiments on the nuclear fuel in extreme conditions, theoretical predictions are if major interest. In a first step, we report comprehensive investigations on point defects in non-stoichiometric uranium dioxide, UO2±x and as a function of temperature using the GGA+U method based on density functional theory. The main result is the predominance of oxygen defects over uranium ones at any composition, due to the correct treatment of the localized 5f electrons. In a second step, such calculations are extended to the determination of the solution and migration energies of some FPs (Ba, Zr, and Cs) in UO2±x. The solubility of these FPs is found to be strongly dependent on their physico-chemical properties and more particularly their availability to form complex oxide phases. Finally, we discuss how such theoretical approach may help in providing insight on the diffusion behaviour of these FPs in UO2±x. References:Phil. Mag. 87, 2561 (2007)Phys. Rev. B 77, 184110 (2008)J. of Nuclear Materials, accepted
4:30 PM - R10.4
Resolution of Fission Gas from Bubbles in Uranium Dioxide.
David Parfitt 1 , Robin Grimes 1
1 Materials, Imperial College, London United Kingdom
Show AbstractClassical molecular dynamics simulations, using a set of previously established pair potentials, have been used to predict the minimum energy needed for krypton and xenon atoms to be resolved from faceted fission gas bubbles contained within a perfect uranium dioxide crystal. The absolute minimum energy for resolution is 43.7 eV for krypton and 68.8 eV for xenon atoms, significantly less that the 300 eV value often assumed in fuel modelling. This value is, however, still sufficient to preclude thermal resolution at normal reactor temperatures. The discrepancies between the present and previous resolution energies are due to the increased probability of absorption at specific impact points on the crystal surface; we have mapped out the probability distribution as a function of impact energy for various impact sites and angles. We also discuss the penetration depth and trajectories followed by atoms impacting upon the surface with initial kinetic energies of up to 2.5 keV.
4:45 PM - R10.5
Stochastic Bubble Density Evolution Modelling of Transit Nuclear Fuel Instability.
Ray Stout 1
1 , RhoBetaSigma Affaires, Livermore, California, United States
Show AbstractStochastic analyses models, in terms of bubble density function and functionals, for bubble density evolution and consequential deformation and stress metrics in nuclear fuel materials are applicable and practical when bubble species are: 1). sufficiently dense; 2). distributed randomly in space; and 3). distributed randomly in size and gas content. Fission gas bubble species in nuclear fuels and spent fuel waste forms nominally satisfy these requirements. A generic, spherical bubble species, identified with a composite variable denoted with b, in nuclear fuels is characterized with physical attribute variables of size [radius a], gas content [number of atoms A], their time rates of change [å, Å], and their transport velocity [v] relative to their local material velocity. Thus, variable b denotes the physical attribute set of variables, [a, å, A, Å, v]. The development of stochastic models starts with the definition of a function space; each function, B(x, t; b)dbdx, of which describes the probable number of bubble species per unit spatial volume element dx per unit species element db about a generic point (x, t; b) at time “t”. Given that future reactor designs will operate at higher temperatures and attain higher burn-ups, bubble density in nuclear fuels and porosity swelling strains, as measured by the number of bubbles per size attributes(radius and radius rate) and per fission gas(atom content and atom content rate), will evolve significantly different from bubbles densities observed in the existing light-water reactors. Models for quasi-steady stochastic evolution of bubble density and bubble density dependent fuel deformation have been formulated for nuclear fuel response models1 in terms of: (1). A bubble density field equation to describe the time evolution for a distribution function of 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 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. A numerical integration of these equations will be discussed to illustrate the response of idealized traction free nuclear fuels’ at different burn-ups during a prescribed temperature excursion. 1. Stout, R.B. [2006]; Bubble Density Dependent Functionals to Describe Deformation and Stress Equilibrium Evolution for In-Reactor Nuclear Fuel Materials, MRS F06, Sym JJ Proc., Paper# 0981-JJ07-04.
5:00 PM - R10.6
Thermodynamic Properties of Actinide-Oxide Solid-Solutions Relevant for Advanced Nuclear Fuels.
Lindsay Shuller 1 , Niravun Pavenayotin 2 , Rodney Ewing 1 3 , Udo Becker 3
1 Materials Science and Engineering, University of Michigan, Ann Arbor, Michigan, United States, 2 Nuclear Engineering and Radiological Sciences, University of Michigan, Ann Arbor, Michigan, United States, 3 Geological Sciences, University of Michigan, Ann Arbor, Michigan, United States
Show Abstract5:15 PM - R10.7
Thermo-mechanical Modeling of Fuel Particle Designs for High-burnup Efficiency in Pebble-bed Nuclear Reactors.
Paul DeMange 1 , Jaime Marian 1 , Alfredo Caro 1 , Magdalena Serrano de Caro 1
1 , Lawrence Livermore National Laboratory, Livermore, California, United States
Show AbstractWith the world’s growing need for energy, interest in nuclear power from High Temperature Gas Reactors (HTGR) has re-emerged in several countries. Interest in HTGR development, in particular, is motivated by several factors, the most important of which include a relative high level of safety achieved by the ability of the fuel system to retain important fission products even at elevated temperatures as well as increased thermal efficiency over Light Water Reactor (LWR) designs. A specific Modular Pebble-Bed Reactor (MPBR) design is being investigated in which the coated fuel particles that are ~1 mm in diameter are embedded in larger graphite pebbles (60 mm in diameter). Each pebble contains roughly ten thousand coated fuel particles while a typical reactor core can contain several hundred thousand pebbles. The pebble design provides additional advantages over other reactor types such as LWRs in that the pebbles are allowed to move through the core during operation. In addition, partially burned and fresh fuel may be added to the top of the core, and pebbles can be checked for burnup and integrity. Fuel particle design is vital to successful HTGR operation. The standard design has been considered as the composition of three different material layers outside the fuel that work in concert to sustain the high pressure and temperature, irradiation, and anisotropic effects such as thermal gradients and layer asphericity without cracking and leakage of the fission products. Ideally, a particle lifetime of several decades is required to allow deep burnup of the kernel fuel, for high efficiency and minimal waste disposal. As such, the number of layers as well as their material makeup and dimensions are the primary considerations in developing a fuel particle design with optimized resistance to failure mechanisms for the desired operating conditions.In this work, a finite element engineering hydrocode is used to perform simulations of various particle designs. The stress behavior of the layers is assessed for fuel particles subject to reactor operating conditions by including detailed models for the equation of state and constitutive and thermal properties of the materials. The objective is to develop a complete fuel particle mechanical modeling platform as well as an optimal fuel particle design capable of long-lifetime sustainability over the burnup of the fuel.