Gianguido Baldinozzi, CNRS-Ecole Centrale Paris
Maria Jose Caturla, Universidad de Alicante
Chaitanya S. "Deo", Georgia Institute of Technology
Chu-Chun Fu, "CEA-Saclay DEN/DMN/SRMP"
Kazuhiro Yasuda, Kyushu University
Yanwen Zhang, Oak Ridge National Laboratory
HH2: Nuclear Fuels I
Monday PM, November 26, 2012
Hynes, Level 1, Room 102
2:30 AM - *HH2.01
Density Functional Theory Calculations of UO2 Oxidation and Diffusion of Fission Gases in UO2plusmn;x
David Andersson 1 Gianguido Baldinozzi 3 Lionel Desgranges 2 Steve Conradson 1 Michael Tonks 4 Paul Millett 4 Blas Uberuaga 1 Chris Stanek 1
1Los Alamos National Laboratory Los Alamos USA2CEA/DEN/DEC, Centre de Cadarache Saint-Paul-lez-Durance France3CNRS-Ecole Centrale Paris Chatenay-Malabry France4Idaho National Laboratory Idaho Falls USAShow Abstract
In this talk we discuss two topics. The first one is oxidation of UO2 and the second one is diffusion of fission gases in UO2±x and its implications for fission gas release models. Formation of hyperstoichiometric uranium dioxide compounds, UO2+x, derived from the fluorite structure was investigated by density functional theory (DFT) calculations. Oxidation was modeled by adding oxygen atoms to UO2 fluorite supercells. A similar approach was applied for studying reduction of U3O8. In agreement with the experimental phase diagram we identify stable line compounds at the U4O9-y and U3O7 stoichiometries. Additionally, we also found a new compound of the U3O7.3333 stoichiometry to be stable between U3O7 and U3O8. The calculated low-temperature phase diagram indicates that the fluorite-derived compounds are favored up to the UO2.5, i.e. as long as the charge-compensation for adding oxygen atoms occurs via formation of U5+ ions. Once U6+ ions are required to achieve overall charge neutrality, the U3O8-y phase becomes more stable. According to our calculations the most stable fluorite UO2+x phases at low temperature (0 K) are based on split quad-interstitial oxygen clusters. This cluster contains four excess and two displaced regular fluorite oxygen ions. It shares some features with the cuboctahedral cluster that is used in existing crystallographic models of U4O9 and U3O7, but the details are different. In order to better understand these discrepancies, the new structure models obtained from our simulations are analyzed in terms of existing neutron diffraction data. Finally, we discuss the importance of cluster formation for oxygen diffusion in UO2+x. In order to better understand bulk fission gas behavior in UO2±x we calculate the relevant activation energies using DFT techniques. Here we focus on Xe, since it is the most important fission gas in UO2 nuclear fuels. By analyzing a combination of Xe solution thermodynamics, migration barriers and the interaction of dissolved Xe atoms with U vacancies, we demonstrate that Xe diffusion predominantly occurs via a vacancy-mediated mechanism. Next we investigate Xe transport on the (111) UO2 surface, which is motivated by the formation of small voids partially filled with fission gas atoms (bubbles) in UO2 under irradiation. Surface diffusion could be the rate-limiting step for diffusion of such bubbles, which is an alternative mechanism for mass transport in these materials. As expected, the activation energy for surface diffusion is significantly lower than for bulk transport. These results are further discussed in terms of engineering-scale fission gas release models.
3:00 AM - HH2.02
Thermodynamics and Kinetics of Defect Disorder in UO2
Abdel-Rahman Hassan 1 X. M. Bai 2 Anter El-Azab 1
1Purdue University West Lafayette USA2Idaho National Laboratory Idaho Falls USAShow Abstract
The microstructure of UO2 is altered significantly by irradiation during its operation as nuclear fuel. Being an oxide, defect disorder, vis-agrave;-vis off-stoichiometry, in this important material is a major aspect of the microstructure change process under irradiation. In general, defect disorder is controlled by the thermodynamic state variables, i.e. temperature and stress, as well as the composition or the oxygen partial pressure in the surrounding environment. We investigate the off-stoichiometric variation in UO2 in terms of atomic defects and electronic carriers. A defect model that describes the defect disorder in terms of temperature and oxygen partial pressure is developed. The underlying parameters are drawn from density-functional theory literature. We extend the model to include the stress and electrostatic fields. Such extension relates the spatial off-stoichiometric variation to the microstructure elements in the oxide. In this regard, the impact of free surfaces and voids forming inside the bulk of UO2 are investigated. The details of the surface interaction with the environment are understood using the ionosorption theory. The results establish the ground for the study of the oxide's time response to transients and irradiation conditions, thus employing the inventory of results from atomistic methods for a better understanding of the oxides structural behaviour. This research was supported as a part of the Energy Frontier Research Center for Materials Science of Nuclear Fuel funded by the U.S. Department of Energy, Office of Basic Energy Sciences under award number FWP 1356, through subcontract number 00122223 at Purdue University.
3:15 AM - HH2.03
Comprehensive Study on Helium Behaviour in Defected UO2 Systems
Zeynep Talip 1 Thierry Wiss 1 Arne Janssen 1 Jean Yves Colle 1 Rudy Konings 1 Joseph Somers 2
1Institute for Transuranium Elements Karlsruhe Germany2Institute for Transuranium Elements Karlsruhe GermanyShow Abstract
Understanding the long term behaviour of the UO2 spent fuel in terms of alpha radiation damage and oxidation is a very important issue for the safety aspects of storage or disposal. Actinides contained in the spent fuel being alpha emitters, often with a long half life, will generate large quantities of helium and possibly cause chemical and physical modifications of the spent fuel matrix. Furthermore, many properties at the atomic scale like defect mobility and self diffusion are strongly dependent on the stoichiometry of UO2. Although helium behaviour in stoichiometric UO2 has been studied since the 1960s, there is still a lack of experimental studies for helium in the fcc lattice of non stoichiometric UO2. In order to assess the effect of the stoichiometry on helium solubility, infusion experiments were performed on hyper and also on hypo-stoichiometric UO2 samples to provide a comprehensive picture of helium behaviour in non-stoichiometric UO2 matrix. Since the UO2-x structure exists only at high temperature, La-doped UO2 samples with various La-content were used. Supplementary work on the oxidation behaviour of UO2 and changes in its local structure were studied by various spectroscopic and microscopy techniques. Helium behaviour was also investigated in 0.1 wt. % 238Pu-doped UO2 samples. The radiation damage build-up and recovery processes were investigated together with the helium behaviour to better understand the possible state of the spent fuel after long storage times. This study will thus give new insights into the relevant aspects to consider for the behaviour of helium in stoichiometric UO2, and defected systems such as nonstoichiometric UO2 and Pu-doped UO2.
3:30 AM - HH2.04
Influence of Grain Orientation on Radiation Induced Strains in UO2 Polycrystals
Philippe Goudeau 1 Etienne Castelier 2 Herve Palancher 2 Axel Richard 2 Jean-Sebastien Micha 3
1Prime Institut Futuroscope France2CEA Centre de Cadarache France3CEA Grenoble FranceShow Abstract
Light ion implantations have generated a lot of interest over the years since they have major technological applications. In nuclear materials studies, they offer the prospect of understanding radiation effects in detail or developing new materials with enhanced radiation resistance properties. Indeed without using costly remote handling and characterization facilities, ion implantation techniques enable the study of effects resulting from neutron irradiations that make samples highly active. The primary effect of loading the surface of a material with foreign elements is to generate swelling of the crystal structure. However, the sample is generally not bulk irradiated but presents an implanted layer the thickness of which typically ranges between a few nanometers and a few microns. The question of how to relate expected swelling in a bulk or surface irradiated sample is therefore essential and we discuss here the first step towards understanding this relationship. Characterization of this swelling effect is usually performed using monochromatic high resolution X-Ray diffraction. However, it does not enable a comprehensive characterization of the strain field in the surface layer loaded with foreign elements for polycrystals. Also, the mechanical models adopted to interpret experiments are usually either simplified (eg. isotropic model) or apply to simplified situations (eg. textured materials) which fails to highlight the more general case in which grain orientation has a major contribution. As a consequence both extensive characterization and accurate modeling of the mechanical state of the implanted layer are required. In this communication, the selected characterization technique (micro-XRD in Laue mode) is first shown to be an efficient method to obtain the strain tensor in the implanted layer at several points within each grain of the polycrystalline samples. Then the strain tensor is demonstrated to be strongly dependent upon crystal orientation. Finally an anisotropic elastic mechanical model involving a free swelling is used to rationalize all the experimental data.
3:45 AM - HH2.05
Densification and Microstructure Characterization of UO2 Processed by Spark Plasma Sintering
Lihao Ge 1 Ghatu Subhash 1 James Tulenko 2 Ronald Baney 2 Sunghwan Yeo 2 Andrew Cartas 2
1University of Florida Gainesville USA2University of Florida Gainesville USAShow Abstract
Conventional sintering of UO2 requires high sintering temperature at around 1700oC in a hydrogen atmosphere for several hours. However, this method requires long production cycle and consumes large amount of energy. Spark plasma sintering (SPS), also known as field assisted sintering technique (FAST), has now become a popular sintering method in various fields due to its merits including rapid processing and low energy consumption. However, very few reports are available in literature on sintering UO2 by SPS. Thus, in this manuscript, a systematic investigation of densification behavior and microstructure of UO2 sintered by SPS is presented. Urania powder of particle size 2.3 microns and grain size 100 nm was used in a graphite die in the SPS machine. Three major sintering parameters (maximum sintering temperature, heating rate and hold time) were varied to investigate their effect on densification, microstructure evolution, Vickers microhardness and Young&’s modulus. The result revealed that the major densification range for UO2 is between 720oC to 1000oC. The sintered material achieved a 96% theoretical density at a maximum sintering temperature above 1050 oC with heating rate of 200oC/min and a hold time of 30 seconds. The entire sintering run duration was around 10 minutes. The effect of hold time and heating rate on final density is different depending on the range of the maximum sintering temperature. At low-temperatures around 850oC, increasing the hold time to 20 minutes and the lowing heating rate to 50 oC increase the final density more significantly than sintering at high temperature range above 1350 oC. The average grain size is increased with the maximum sintering temperature and the hold time. There was no significant effect of the heating rate on the average grain size. Typical grain size varied from 2 microns to 6 microns depending on the processing conditions. The mechanical properties of the SPS sintered pellets have been measured and compared with the value reported in the literature. The average hardness of the pellets is around 6.4±0.4GPa and shows a Hall-Petch correlation with the average grain size. The measured Young&’s modulus using ultrasonic measurements is 204±18GPa and it increased with the theoretical density. Both of these results are in agreement with literature for UO2 produced by conventional methods.
HH3: Radiation Effects I - Microstructures
Monday PM, November 26, 2012
Hynes, Level 1, Room 102
4:30 AM - *HH3.01
An Attempt to Handle the Nanopatterning of Materials Created under Ion Beam Mixing
Simeone David 1 2 Laurence Luneville 3 2 Baldinozzi Gianguido 2 1
1CEA Saclay France2CNRS Chatenay Malabry France3CEA Saclay FranceShow Abstract
Nanocomposite materials provide many opportunities for synthesizing materials with improved or unique properties. The challenge for exploiting this potential resides in the difficulty in elaborating such materials at the nanometric scale. An attractive way to overcome this difficulty is to employ the self organization of the composition field generated by ion beam mixing at the atomic scale. This composition field is mainly driven by thermal spikes induced by the slowing down of incident particles. Despite some attempts were made to describe the effect of a thermal spike on the composition field, no clear analysis of the effect of a displacement cascade on this field has been established. Using a simple Cahn Hiliard model, we present a first attempt to describe this effect. In this presentation, we focus our attention on the concept of “effective temperature” to describe the ion beam mixing.
5:00 AM - *HH3.02
Computational Modeling of Irradiation Effects in Nuclear Structural and Fuel Materials
Kazunori Morishita 1 Yasunori Yamamoto 2 Yoshiyuki Watanabe 3
1Kyoto University Uji Japan2Kyoto University Uji Japan3Japan Atomic Energy Agency Rokkasho JapanShow Abstract
In nuclear energy systems, structural and fuel materials suffer from high energy neutron bombardment, which causes material&’s microstructure changes, resulting in the degradation of materials&’ performance. This may lead to serious problems in the integrity of reactor performance. Not only for development of superior radiation-resistant materials but also for establishment of the methodology of efficient and effective reactor maintenance, the degradation of material&’s performance due to irradiation should be well understood, enough monitored, precisely predicted, sufficiently controlled, and reasonably suppressed, on a basis of reliable radiation damage physics. In the present study, multiscale radiation damage processes were theoretically investigated using such several computational modeling methods as ab-initio calculations, molecular dynamics (MD) calculations, kinetic Monte-Carlo (KMC) calculations, and kinetic rate theory based calculations. Nucleation and growth of point defect clusters in nuclear metals during irradiation was simulated using KMC and rate equation calculations, in which defect energetics was obtained by ab-initio and MD calculations. Simulated microstructural evolution in irradiated metals was obtained as a function of dpa/s. It shows that the nucleation and growth of voids depends much on dpa/s, while that of SIA loops does not. Also, in order to understand the oxidation process of Zr alloy, the fuel cladding material in commercial light-water reactor, the migration behavior of an oxygen atom in ZrO2 was investigated by ab-initio calculations and diffusion equation analysis as a function of the stress and temperature gradient applied to ZrO2. Our results indicate that the oxidation rate of Zr is in approximately proportion to cubic root of time, which is consistent with experiments.
5:30 AM - HH3.03
Modeling and Simulation of Restructuring in Irradiated Materials by Pore and Grain Boundary Migration
Michael R Tonks 1 Liangzhe Zhang 1 Paul Millett 1 Xianming Bai 1 Bulent Biner 1
1Idaho National Laboratory Idaho Falls USAShow Abstract
The grain size in nuclear fuel has a large effect on critical phenomena such as fission gas release and creep. Thus, our ability to predict these important phenomena will be improved by a better understanding of grain growth under irradiation. In this study, a phase field model is used to study the interaction between pore and grain boundary migration under a high temperature gradient. We begin with a detailed study of the interaction between a single pore and a bicrystal GB. We then do a more complicated simulation of many pores migrating in a polycrystal. We also discuss how this information will be used to develop a model of the average grain size as a function of temperature for a fuel performance code.
5:45 AM - HH3.04
Nanostructuration of Cr/Si Layers Induced by Ion Beam Mixing
Laurence Luneville 1 Ludovic Largeau 2 Cyrile Deranlot 3 Nathalie Moncoffre 4 Yves Serruys 1 Gianguido Baldinozzi 5 David Simeone 1
1CEA Gif sur Yvette France2CNRS Marcoussis France3CNRS/Thales Palaiseau France4IN2P3 Lyon France5CNRS/ECP Chatenay-Malabry FranceShow Abstract
Atomic collisions in solids induced by ion beam are often associated with the concept of disorder. In fact, the mobility induced in solids by ion irradiation at appropriate temperatures leads to the production of a wealth of phases which may (or not) be related to the equilibrium phase diagram. Despite, many attempts are made to understand the phase stability and the enhanced mobility of defect under irradiation even at the atomic scale, no clear picture of ion beam mixing exists. The major problem associated with ion beam mixing comes from the fact that it remains quite difficult to accurately measure a concentration over few nanometers. The X Ray Reflectometry (XRR), extensively used in micro electronics, appears to be a useful technique to overcome this difficulty. In this work, we apply the XRR technique to study the nanostructuration of Cr/Si layers induced by 80 keV Kr ion beam at room temperature, a textbook example of ion beam mixing. The analysis of XRR profiles allows computing accurate profiles of Si and Cr concentrations with a resolution equals to 1 nanometer. From these experimental profiles, we point out that the ion beam mixing appears to be a complex process which can not be only described as a diffusion controlled process.
HH1: Steels -- From Point Defects to Mechanical Properties
Monday AM, November 26, 2012
Hynes, Level 1, Room 102
9:30 AM - *HH1.01
Pathways for Development of Advanced Materials for Nuclear Energy Systems
Steven John Zinkle 1 Michael P. Brady 2 Bruce A. Pint 2 Lance L. Snead 2 Lizhen Tan 2 Kurt A. Terrani 1 Yuki Yamamoto 2
1Oak Ridge National Lab Oak Ridge USA2Oak Ridge National Lab Oak Ridge USAShow Abstract
This presentation will review some of the current and emerging strategies to develop high-performance materials with simultaneous high radiation resistance, high strength, good toughness and corrosion resistance, and moderate fabrication cost. There are three general approaches for designing radiation resistance: Nanoscale precipitates or interfaces to produce high point defect sink strength; purposeful utilization of immobile vacancies; and utilization of radiation-resilient matrix phases. In the future, utilization of advanced manufacturing processes to produce near-net shape parts with precise microstructural control will be of increasing importance to control fabrication costs and to create high-performance fabrication architectures that could not be achieved using conventional fabrication methods. Recent progress on development of high-performance steels designed using computational thermodynamics will be summarized. These steels are designed to incorporate a high density of highly stable nanoscale precipitates that could serve as efficient point defect recombination centers during irradiation, and also provide good thermal creep strength at high temperatures. Some of the options under consideration for potential accident tolerant cladding and fuel systems for commercial fission light water reactors will be discussed. Desirable attributes for the cladding under postulated loss of coolant accident and station black out conditions include oxidation resistance to steam and air and good thermal creep strength at temperatures in excess of 1200oC, along with standard cladding requirements for fabricability, low parasitic neutron absorption, hermetic containment of fission products, good compatibility with the fuel and fission products, and acceptable cost.
10:00 AM - HH1.02
Atomic Kinetic Monte Carlo Modeling of Concentrated FeCrNi Alloys Based on ab initio Calculations
Christophe Domain 1 2 Charlotte S Becquart 3 2 Jean Baptiste Piochaud 3 2
1EDF Ramp;D Moret sur Loing France2EM2VM Paris France3UMET Lille FranceShow Abstract
Internal structure of pressurized water reactors are made of austenitic materials. Under irradiation, the microstructure of these concentrated alloys evolves and solute segregation on grain boundaries or irradiation defect such as dislocation loops are observed to form. In order to model and predict the microstructure evolution, a multiscale modeling approach needs to be developed, which starts at the atomic scale. Atomic Kinetic Monte Carlo (AKMC) modeling is the method we chose to provide an insight on defect mediated diffusion under irradiation. In that approach, we model the concentrated commercial steel as a FeCrNi alloy (γ-Fe70Cr20Ni10). As no reliable empirical potential exists at the moment to reproduce faithfully the phase diagram and the interactions of the elements and point defects, we have adjusted a pair interaction model on DFT calculations. The point defect properties in the Fe70Cr20Ni10, and more precisely, how their formation energy depends on the local environment will be presented and some AKMC results on irradiation defect formation and solute segregation will be presented.
10:15 AM - HH1.03
Molecular Dynamics Simulations of Cascade Evolution near Trapped Interstitial Clusters
Nathan Capps 1 Aaron Kohnert 2 Karl Hammond 1 Xu Donghua 1 Brian Wirth 1
1University of Tennessee Knoxville USA2University of California Berkeley Berkeley USAShow Abstract
The overlap of displacement cascade is believed to be important in the development of visual defect clusters in thin film, in-situ ion irradiation studies. In this work, we use molecular dynamics simulations to investigate how impurities and damage induced by displacement cascades affect the mobility of a pre-existing interstitial-type dislocation loop in BCC iron. It is well known that impurities, such as helium, carbon, and nitrogen affect the ability of interstitial dislocation loops, and are likely responsible for difference in loop diffusivities between computer simulations and experimental observations by transmission electron microscopy. We have used molecular dynamics simulations to evaluate whether a displacement cascade could result in the de-trapping of an interstitial cluster from interstitial impurity atoms. By varying the energy and directional velocity of the primary knock on atom (PKA), we are able to observe how the trapped defect reacts with the cascade damage. Our initial simulation results reveal that cascades caused PKAs of energy greater than 10 KeV can cause the loop to de-trap from impurities, but that it may rapidly become trapped in the cascade debris. Furthermore, on several occasions, the cascade has induced a change in orientation or Burgers vector of the dislocation loop. The presentation will summarize the molecular dynamics simulation results as a function of PKA energy, distance from the trapped loop and direction, as well as the effect of loop size on the probability for de-trapping and subsequent diffusion. These simulation results will be used to inform cluster dynamics models of dislocation loop evolution in irradiated ferritic/martensitic alloys.
10:30 AM - HH1.04
Atomistic Simulations of Decomposition Kinetics in Fe-Cr Alloys: Influence of the Magnetism
Oriane Senninger 1 Enrique Martinez 2 Frederic Soisson 1 Chu Chun Fu 1
1CEA Gif sur Yvette France2Los Alamos National Laboratory Los Alamos USAShow Abstract
Ferritic Stainless steels are commonly used as structure materials in the nuclear industry. The prediction of the lifetime of current power plans and the development of new generations of nuclear reactors raise the question of the possible decomposition of iron-chromium alloys. Atomistic simulations are especially useful to study the kinetics of decomposition during thermal ageing and under irradiation. We propose here a modeling of precipitation in binary iron-chromium alloys during thermal ageing by Atomistic Kinetic Monte Carlo simulations in a rigid lattice approximation. The system evolves by vacancy diffusion. Pair interaction energies are fitted on the thermodynamic and diffusion properties of the alloy. Both of these properties are highly influenced by the magnetic evolution of the alloy with the concentration of chromium and the temperature. On the one hand, the mixing energy of the alloy faces a change of sign according to the concentration which lead to an asymmetrical phase diagram. To take this characteristic into account, a dependency on the local chromium concentration is introduced in interaction energies. By this fit, we obtain solubility limits in good agreement with most recent review studies. On the other hand, the ferro to paramagnetic transition strongly accelerates the diffusion near the Curie temperature. As this evolution affects the kinetic evolution of the alloy, we reproduce this increase of the diffusion coefficients by introducing a correcting parameter on the migration barriers depending on the local concentration and the temperature. In order to evaluate our model, we compare the Monte Carlo simulations with existing Small Angle Neutron Scattering kinetic experiments at various compositions and temperatures. Some of these experiments are in the ferromagnetic configuration, others in the paramagnetic one. We obtain a good agreement between experiments and simulations for both magnetic configurations.
11:30 AM - HH1.06
Distribution of Void Swelling and Irradiation Creep and Resultant Strains in Thick 304 Stainless Steel Hexagonal Reflector Blocks in Response to Spatial Gradients in Neutron Flux-spectra and Irradiation Temperature in EBR-II
Frank A. Garner 1 Paula D. Freyer 2 Douglas L. Porter 3 James Wiest 3 Collin J. Knight 3 M. Sagisaka 4 Y. Isobe 4 J. Etoh 4 T. Matsunaga 4 T. Okita 5 Yina Huang 6 Jorg Wiezorek 7
1Radiation Effects Consulting Richland USA2Westinghouse Electric Company Pittsburgh USA3Idaho National Laboratory Idaho Falls USA4Nuclear Fuel Industries Osaka Japan5University of Tokyo Tokyo Japan6University of Wisconsin Madison USA7University of Pittsburgh Pittsburgh USAShow Abstract
Void swelling and irradiation creep can be life-limiting processes for components of both fast reactors and light-water reactors. Predictive equations are therefore required to forecast safety and economic limits of operation. It is not generally recognized, however, that such equations are almost always developed from rather limited numbers of specimens, all of thickness 0.3-1mm. In such configuration there are no significant variations in temperature, stress or neutron flux-spectra across the specimen thickness. However, in actual reactor components, especially in PWR internals or fast reactor reflector assemblies, thicknesses can be larger with significant gradients not only in environmental variables (temperature, dpa rate, stress) but in the resulting distribution of macroscopic stresses and strains. There are currently no benchmark data fields that allow confident incorporation of such "thin" equations in design codes for "thick" and complex shapes, especially since swelling-creep interactions determine the local stress field and feed back into both swelling and creep strains. We have examined a series of five hexagonal cross-section reflector blocks (annealed 304SS, 50mm flat-to-flat, ~250mm length) that were vertically stacked in a thin-wall (1mm) hexagonal 304SS wrapper can in Row 8 of EBR-II in flowing sodium. During their residence in core the blocks accumulated 0.5-33 dpa depending on their axial position in the assembly. Over the stack there were significant axial and radial gradients in both dose and temperature with gamma heating leading to significant internal temperature increases, producing a complex spatial distribution of void swelling and creep strains. Four of these blocks have been subjected to non-destructive examination, and two of these to extensive destructive examination thereafter. Measurements involved profilometry, ultrasonic time-of-flight measurements to map the internal distribution of swelling, backed up by density change measurements and electron microscopy. Asymmetrical internal peaks approaching ~4% swelling were found in the center of the blocks with average swelling of ~1.5% over the block cross section. It was found that the complex internal distribution of microscopic strains arising from void swelling and carbide densification can be related to the macroscopic deformation of the blocks, producing both bulging and bowing of the blocks. The strains of the blocks could also be correlated with the measured strains of the hexagonal can that housed them. These data not only provide unique insights on the interrelationships between swelling and irradiation creep but can also serve as a benchmark for computer code calibration for prediction of distortion of thick components.
11:45 AM - HH1.07
Effect of Radiation on Embrittlement and Matrix Cu Content of a RPV Weld with Different PWHT Conditions
Mikhail A. Sokolov 1 Randy K Nanstad 1 Michael K Miller 1 Ken Littrell 1
1ORNL Oak Ridge USAShow Abstract
The influence of temperature, hold time, and cooling rate on the Charpy impact properties and the copper level in the matrix has been investigated on a weld fabricated from the same weld wire used for HSSI Weld 73W and a Linde 80 flux before and after irradiation to 0.8x1019 neutron/cm2 (E>1MeV). This weld has a relatively high bulk copper content, 0.32% wt, in as-welded condition. The heat treatment consisted of heating the material to the desired temperature, holding at the post-weld heat treatment (PWHT) temperature, and then cooling down to room temperature. Except for special cases, all PWHTs were performed with a heating and cooling rate of 15oF/h (8oC/h) to simulate the heating/cooling rate of a real vessel. In two special cases, material was heated with 15oF/h (8oC/h) rate but water quenched after holding at the PWHT temperature. The highest PWHT temperature was 650oC/24h, while the other PWHTs were 610oC/24h, typical PWHT of reactor pressure vessels (RPV), and 580oC/100h. In addition, a part of the material after 610oC/24h PWHT was heat treated at 454oC/168h to simulate a post-irradiation annealing. Charpy impact properties were measured using sub-size 3x4 mm specimens and matrix Cu content was measured by atom probe tomography before and after irradiation. Small-angle neutron scattering was used to measure number densities and size of copper-rich precipitates in the irradiated specimens. Charpy specimens were irradiated in the Ford Reactor at 288oC. It was found that the higher PWHT temperature resulted in higher Charpy upper-shelf energy (USE) with little effect on the ductile-to-brittle transition temperature (DBTT). The lower PWHT temperature and slower cooling rate were found to be beneficial in reducing the matrix Cu content. The matrix Cu content after irradiation to 0.8x1019 neutron/cm2 was approximately the same for all three welds measured regardless of their different matrix Cu contents in the unirradiated condition. Consequently, the weld with the lowest PWHT temperature (and lowest matrix Cu) exhibited the lowest shift of DBTT and drop in USE. These results are proving the postulate about impotence of copper in solution (matrix Cu) rather than bulk Cu content in radiation embrittlement of RPV materials. Additional annealing at 454oC/168h after 610oC/24h PWHT did not show any additional effects on subsequent radiation embrittlement. Atom probe tomography (MKM) sponsored by the Scientific User Facilities Division (ORNL ShaRE User Facility), Office of Basic Energy Sciences, U.S. Department of Energy.
12:00 PM - HH1.08
Hardening and Softening Caused by Long Term Neutron Irradiation in Modified-SUS316 Stainless Steel
Hiroshi Oka 1 Tomoki Kubota 1 Naoyuki Hashimoto 1 Somei Ohnuki 1 Shinichiro Yamashita 2
1Hokkaido University Sapporo Japan2JAEA Ibaraki JapanShow Abstract
To estimate irradiation effects such as irradiation hardening or softening both in fission and fusion reactor components, an appropriate equation describing relationship between microstructure and mechanical properties in actual length scale is absolutely required. On the other hand, austenitic stainless steels have been used extensively in fission reactor applications, so that the most extensive database for nuclear applications would be based on the steels. In this study, modified-SUS316 austenitic stainless steels irradiated in a fast reactor were investigated to construct an equation connecting their micro- and macro-structures and mechanical properties with utilizing TEM, tensile test and hardness test. A modified-SUS316 stainless steel (PNC316), which has been developed as a cladding tube material with superior high temperature strength and swelling resistance for a liquid metal cooled fast reactor, was examined in this study. The representative chemical composition of PNC316 is Fe-16Cr-14Ni-2.5Mo-0.25P-0.004B-0.1Ti-0.1Nb. It is, in most cases, used in the 20% cold-worked condition. Neutron irradiation was performed using the experimental fast reactor JOYO. The range of irradiation temperature and dose were 750-1000 K and 16-103 dpa corresponding to irradiation time from 4,000 to 20,000 h, respectively. Correlations of yield strength and Vickers-hardness with irradiation-induced microstructure were investigated. Irradiation softening in yield strength occurred in whole temperature range in the present study, probably due to a recovery of existing dislocation structure. Large precipitates like Laves phase, M6C and/or M23C6 were observed by TEM, especially in high temperature irradiation condition. The estimation of radiation-induced change in yield strength, Δσy, based on barrier hardening model developed by short-range and long-range obstacles showed small value compared to actual Δσy. It is assumed that concentration of carbon in the matrix was decreased due to precipitation of irradiation-induced large carbides, resulting in a loss of solution hardening.
12:15 PM - HH1.09
Microstructural Characterization of Activated Materials with Neutron and X-Ray Diffraction
Donald Brown 1 Thomas Sisneros 1 Bjorn Clausen 1 Levente Balogh 1 Jon Almer 2
1Los Alamos National Lab Los Alamos USA2Argonne National Lab Argonne USAShow Abstract
Diffraction is well suited to the characterization of microstructures. Neutron and high energy x-ray diffraction, in particular, have some undeniable advantages for studying activated samples. Neutrons and high energy x-rays penetrate millimeters into most materials, providing a statistically relevant probe of the microstructure in the bulk of the material. Moreover, little or no hazardous and costly sample preparation is necessary, which enables repeat tests on the same sample or even in-situ measurements under simulated operating conditions. Finally, neutron detectors are often insensitive to background gamma radiation emitted from activated samples. The SMARTS diffractometer at the Lujan Center was designed to study engineering materials under their operating conditions and, as such, has sophisticated sample environments enabling in-situ ND studies during deformation and at non-ambient temperatures. This talk will use the example of in-situ diffraction measurements during annealing and deformation of irradiated HT-9 steel to highlight the capabilities of the instrument, in particular, related to the study of activated materials.
12:30 PM - HH1.10
Crystal Plasticity Modeling of Localized Deformation in Irradiated bcc Metals
Anirban Patra 1 David L. McDowell 1 2
1Georgia Institute of Technology Atlanta USA2Georgia Institute of Technology Atlanta USAShow Abstract
Structural materials used in nuclear environments are subjected to significant doses of radiation at elevated temperatures. This gives rise to a large number of point defect clusters, dislocation loops, and complex dislocation networks in the irradiated metals. Irradiation-induced defects hinder the glide of mobile dislocations (that carry inelastic deformation) through the material. An increase in yield strength, followed by flow localization and lower engineering strain to fracture is observed during quasi-static tensile loading. The post-deformation microstructure reveals that the inelastic strain is localized along narrow dislocation channels which are ‘cleared&’ of majority of the irradiation-induced defects. The present work uses a continuum constitutive crystal plasticity framework to model the mechanical behavior and deformed microstructure in irradiated bcc metals, with emphasis on capturing the aforementioned localization phenomena. Material defects generated due to irradiation (interstitial loops in bcc) and dislocations (mobile and immobile) are used as substructure variables in this model. The substructure evolution equations are based on physical mechanisms of dislocation-dislocation and dislocation-defect interactions. The framework is used to simulate the localization phenomena for a model ferritic/martensitic steel subjected to irradiation and then loaded in tension. Events leading to flow localization along the dislocation channels are studied. Qualitative effects of varying the grain size, initial crystallographic orientations, and degree of cross-slip on the localization behavior are also studied.
Gianguido Baldinozzi, CNRS-Ecole Centrale Paris
Maria Jose Caturla, Universidad de Alicante
Chaitanya S. "Deo", Georgia Institute of Technology
Chu-Chun Fu, "CEA-Saclay DEN/DMN/SRMP"
Kazuhiro Yasuda, Kyushu University
Yanwen Zhang, Oak Ridge National Laboratory
HH6: He and H Effects
Tuesday PM, November 27, 2012
Hynes, Level 1, Room 102
2:30 AM - *HH6.01
Synergistic Effects of H and He on Radiation Damage of ODS Ferritic Steel under Triple-ion Irradiation
Luke Hsiung 1 M. Fluss 1 S. Tumey 1 B. Choi 1 E. Meslin 2 Y. Serruys 2 A. Kimura 3
1Lawrence Livermore National Laboratory Livermore USA2CEA Gif-sur-Yvette France3Kyoto University Kyoto JapanShow Abstract
Synergistic effects of H and He on radiation damage of Fe-16Cr-4.5Al-0.3Ti-2W-0.37Y2O3 ODS ferritic steel ODS steel under triple-ion irradiation have been investigated using high-resolution transmission electron microscopy (HRTEM) techniques. The frequent observations of partially crystallized Y4Al2O9, YAlO3, and Y2TiO5 complex-oxide nanoparticles/clusters indicate that the nanoparticles/clusters in as-fabricated MA/ODS steel are not only structurally defective but also chemically deviate from equilibrium. Simultaneous triple-ion-beam consisting of Fe8+, H+, and He+ were employed for irradiation of the ODS steel at 600C. Results of the experiment reveal that a constructive effect of defective nanoparticles/clusters on the suppression of radiation-induced swelling can be revealed through the observations of helium-filled cavities (bubbles) trapped preferentially at the interfaces between the matrix and nanoparticles/clusters. An adverse effect of hydrogen implementation on the triple-ion irradiated ODS ferritic steel is readily observed through the formation of hydroxide compound in association with large facetted voids. The formation of HFe5O8-base hydroxide compound (space group: P63mc) with lattice parameter: a = 0.598 nm and c = 0.937 nm was identified presumably for the first time in triple-beam irradiated ODS ferritic steel. This work was performed under the auspices of the U.S. Department of Energy by Lawrence Livermore National Laboratory under Contract DE-AC52-07NA27344.
3:00 AM - HH6.02
A First-principles Model for Noble Gas Defects in Iron and Non-magnetic BCC Refractory Metals
Duc Nguyen-Manh 1 Sergei L. Dudarev 1
1Culham Centre for Fusion Energy Abingdon United KingdomShow Abstract
Generation of helium in materials through transmutation nuclear reactions under high-energy neutron irradiation, giving rise to radiation swelling and grain boundary embrittlement, is a major factor limiting the lifetime of structural materials in fusion and fission power plants. Chemically, helium is similar to other noble gases, for example neon and argon, and the development of an accurate predictive model for defects formed due to accumulation of helium and other noble gases in the crystal lattice of iron, steels and non-magnetic body-centred cubic (bcc) refractory metals and alloys, offers a way of quantifying the effect of transmutation reactions on the structural integrity of reactor components. So far, experimental and modelling effort has been focused exclusively on helium defects and the combined synergetic effects associated with simultaneous accumulation of helium and hydrogen in materials. It remains unclear to what extent ion-irradiation experiments, involving other noble gas atoms, could be used to explore the radiation-induced phenomena resulting from accumulation of transmutation products in materials. Reliable experimental information on noble gas defects is scarce, and application of first-principles density functional methods is probably the fastest and most accurate presently available method of acquiring quantitative information about noble gas defects. We have carried out a comparative study of defects, resulting from the incorporation of noble-gas atoms (He, Ne, Ar, Kr, Xe) into bcc transition metals, using first-principles density functional theory (DFT) calculations. The formation energies for noble-gas atoms in various substitutional and interstitial configurations have been calculated to understand the trends and quantify the local lattice distortion effects as a function of the noble-gas atom size. Helium is a relatively small atom and He defect energies are lower than those corresponding to other noble gas atoms. The size effect results in the change of the relative stability of tetrahedral and octahedral interstitial sites for Ne, Ar, Kr and Xe in comparison with He. Furthermore, the most stable double-gas-atom configuration changes from the <110> in bcc-Fe and <111> in bcc-W configurations for the He-He case into the <100> one for other noble-gas atoms. We also investigate the stability of bound vacancy-noble-gas configurations which are expected to promote void nucleation under irradiation. Finally, the dependence of vacancy-noble gas binding energies in iron and other bcc transition metals has been investigated as a function of vacancy/noble-gas atom ratio to provide input for kinetic Monte-Carlo simulations of microstructural evolution under irradiation.
3:15 AM - HH6.03
Energetic Landscape and Diffusion of He at Grain Boundaries in bcc-Fe from Atomistic Simulations
Chu-Chun Fu 1 Lei Zhang 1 2 Guang-Hong Lu 2
1CEA Gif sur Yvette France2Beihang University Beijing ChinaShow Abstract
In addition to intrinsic defects, large amounts of He and H are also produced by nuclear transmutation under high energy neutron irradiation. For instance, structural materials of future fusion devices may suffer from swelling, intra- and inter-granular embrittlement due to the accumulation of He. As a first and indispensable step to understand these macroscopic properties, the energetic and mobility of He atoms in bcc iron need to be accurately determined. Both first-principles and classical molecular dynamics methods are employed to investigate the lowest-energy sites and migration mechanisms of He in various grain boundaries (GBs) with different characteristics. We have first optimized the GB structure by adding either vacancies or self-interstitial atoms close to the interfacial plane. Then, He formation energies for all the possible sites are evaluated. The obtained values are indeed much lower (~1.7 eV) near the GBs than in the bulk, indicating a strong He segregation tendency. In addition, isolated He atoms reveal to reduce significantly the GB cohesion, although without forming bubbles. We show that the He formation energies are determined by the interplay of two competing contributions: the available volume for He insertion, and the distortion of the Fe lattice. Also, both 0K and finite temperature migration barriers for He and vacancy diffusion in the various GBs are calculated. Interestingly, an interstitial He always requires a larger energy barrier for diffusion along the GBs than in the bulk, at variance with the case of vacancy migration. The present results are used as input data for larger-scale models in order to study the kinetic of He segregation and bubble formation at GBs.
3:30 AM - HH6.04
He Bubble Formation in bcc Fe and Their Interaction with Radiation
Roger Smith 1 Xiao Gai 1 Steven Kenny 1
1Loughborough University Loughborough United KingdomShow Abstract
The structure of He bubbles in bcc Fe is discussed from a simulation point of view. Formation energies are calculated for different sizes of vacancy-He clusters using classical interaction potentials. Molecular dynamics simulations are then performed to investigate the effect of low energy collision cascades on bubbles of different sizes. The results show that it is possible for both vacancies and Fe interstitials to be attracted to the bubbles depending on their relative size and stability. Some preliminary long time scale dynamics results will also be presented to indicate possible mechanisms by which the He bubbles can form.
3:45 AM - HH6.05
Helium Implantation Effects on the Compressive Response of Cu and Fe Nano-pillars
Qiang Guo 1 Peri Landau 1 Peter Hosemann 2 Julia Greer 1 3
1California Institute of Technology Pasadena USA2University of California at Berkeley Berkeley USA3California Institute of Technology Pasadena USAShow Abstract
Nanomechanical experiments and site-specific microstructural analysis on as-fabricated and helium (He)-implanted 120nm-diameter single crystalline Cu and Fe nano-pillars revealed specific effects of He implantation on the mechanical properties. In particular, the <111>-oriented Cu pillars implanted with 0.35±0.05 at. % He throughout the gauge section were found to yield at 1.2GPa under uniaxial compression. This value is 30% higher than the yield strength of as-fabricated samples with the same diameter. Stress-strain data of the implanted Cu pillars exhibits shorter and more frequent strain bursts, as well as notable strain hardening with a hardening slope of 3.52±0.82GPa. In the case of <110>-oriented Fe pillars, samples implanted with 0.33±0.05 at. % He gave a compressive flow stress of 2.0GPa at 5% strain, a value ~25% higher than the 1.5GPa flow stress of as-fabricated pillars. In both as-fabricated and implanted Fe pillars, the stress-strain behavior was shown to have 3 distinct regimes, starting from elastic loading followed by continuous strain hardening, and eventually reaching a “steady state” where the flow stress remained a constant. Our findings were interpreted in terms of the interaction between mobile dislocations and the implantation-induced defects.
4:00 AM - HH6.06
Spatial Heterogeneity of Interface Energy Stabilizes Sub-nanometer Interfacial He Platelets
Abishek Kashinath 1 Michael J. Demkowicz 1
1Massachusetts Institute of Technology Cambridge USAShow Abstract
Using atomistic modeling, we find that He is trapped at Cu-Nb interfaces in the form of sub-nanometer sized platelet-shaped clusters at misfit dislocation intersections (MDIs). This behavior occurs due to the spatial heterogeneity of Cu-Nb interface energy: He-vacancy clusters wet regions of high interface energy at MDIs while avoiding low interface energy regions. Below a critical interface He concentration, these platelets are stable even in the presence of high vacancy supersaturation. The consequences of this finding for preventing He damage in fusion applications will be discussed. This material is based upon work supported as part of the Center for Materials at Irradiation and Mechanical Extremes, an Energy Frontier Research Center funded by the U.S. Department of Energy, Office of Science, Office of Basic Energy Sciences under Award Number 2008LANL1026.
4:30 AM - HH6.07
The Effect of Helium Bubbles on Hydrogen in Tungsten
Niklas Juslin 1 Faiza Sefta 1 Brian D. Wirth 1
1University of Tennessee Knoxville USAShow Abstract
Tungsten is a candidate material for the divertor in fusion reactors. The divertor will be subject to intense, low energy (1-100 eV) hydrogen isotope and helium bombardment from the plasma. He and H in a material can cause changes in thermal and mechanical properties, such as swelling, ductile to brittle transition temperature, bubble formation and nanofuzz formation. Fuel retention, in particular of tritium, is a serious issue. Molecular dynamics is a valuable tool to study the energetics and structures of H and He clusters and many radiation damage phenomena that happen on short time and length scales, up to nanoseconds and millions of atoms. Using molecular dynamics simulations, we have studied the effect of voids and helium bubbles on hydrogen binding to bubbles and defects, hydrogen retention and hydrogen diffusion. Hydrogen introduced in bubbles of different size and composition quickly diffuse towards the edge of the bubbles, but stay bound around the first atomic layer of the surrounding W matrix. Only at high temperatures, 1500 K and above, a significant amount of hydrogen escapes further than a few Å from the bubble. Helium, on the other hand, stays strongly bound within the bubble. By comparing the energetics and positions of hydrogen defects in the tungsten matrix, near the edge of the bubble and inside the bubble, we note that the edge is about 1 eV more favorable than the ground state position in the matrix. A bubble near the surface can burst, expelling the gas atoms, depending on bubble pressure and distance to surface. As the hydrogen is preferentially bound to the edge of the bubble, a significant amount of the hydrogen is retained in the surface, while the helium is expelled and a crater is formed. We study the effect this has on gas atom composition in the surface and surface morphology. By implanting H and He in surfaces with pre-existing high pressure He-H bubbles close to the surface, we can study both surface roughening and the different roles of hydrogen and helium during plasma exposure.
4:45 AM - HH6.08
Helium Bubbles in Iron: Stability, Equilibrium and Effect to Strengthening
Yury Osetskiy 1 Roger Stoller 1
1ORNL Oak Ridge USAShow Abstract
Helium atoms formed in steels by transmutation reactions under neutron irradiation lead to He-filled bubble formation that affect strongly microstructure evolution and mechanical properties. We present here the results of atomic-scale modeling study of properties of He-filled bubbles in Fe. We have investigated equilibrium state as a function of bubble size, ambient temperature and He content as well as effect of He-bubbles to dislocation motion. The results obtained on bubble stability allow us to formulate an equation of state for He-bubbles that can used in theoretical and computational models for prediction of microstructure evolution in steels under neutron irradiation. The results on bubble-dislocation interaction allow the prediction of mechanical property change in irradiated steels.
5:00 AM - HH6.09
Nanoporous Metals for Prevention of Helium Bubble Formation
Patrick Cappillino 1 Benjamin W Jacobs 2 Michelle A Hekmaty 1 Ryan Hartnett 1 David B Robinson 1
1Sandia National Laboratories Livermore USA2Protochips, Inc. Raleigh USAShow Abstract
Helium-3 buildup is an important mechanism of aging in metal tritides. Accumulation eventually leads to formation of bubbles of 3He gas, causing changes in material properties and, ultimately, uncontrolled release. It has been hypothesized that this problem would be mitigated in nanoporous Pd, since all points in the metal lattice should be near a gas/solid interface, allowing 3He to diffuse harmlessly out of the solid. We have shown that mesoporous Pd and Pd alloy powders can be synthesized in a scalable fashion using soft templates. We have produced batches of mesoporous Pd on the gram scale, with regular arrays of pores having diameters that are tunable between 3 and 13 nm by chemical reduction of Pd2+ salts around hexagonally packed cylindrical micelles of surfactants or block copolymers. This control of pore size is effected by varying the composition and/or size of the molecular template. The pore size has effects hydrogen storage capacity and kinetics, as well as thermal stability. Preliminary results of helium implantation experiments show that helium is not retained in these materials. Sandia National Laboratories is a multi-program laboratory managed and operated by Sandia Corporation, a wholly owned subsidiary of Lockheed Martin Corporation, for the U.S. Department of Energy&’s National Nuclear Security Administration under contract DE-AC04-94AL85000.
5:15 AM - HH6.10
Stopping Power of H and He in Al from Plane-wave Ehrenfest Molecular Dynamics
Andre Schleife 1 Yosuke Kanai 2 1 Alfredo Correa 1
1Lawrence Livermore National Laboratory Livermore USA2The University of North Carolina at Chapel Hill Chapel Hill USAShow Abstract
During the initial stages of the deceleration of a fast atom in a bulk material, interactions of the moving atom with the electronic system of the target material are the dominant mechanism. For several decades the interest in understanding these processes has been growing and has prompted both experimental and theoretical studies, ultimately aiming at producing novel materials with a high radiation tolerance. In recent years predictions from parameter-free computational studies are becoming a useful tool in this context: Predictive results can now be achieved due to the use of real-time time-dependent densityfunctional theory for modeling the interaction of the projectile with the electronic system of the target, overcoming the problematic adiabatic Born-Oppenheimer approximation. Our newly developed first-principles implementation of Ehrenfest dynamics is based on explicit integration of the time-dependent Kohn-Sham equations in real time. The wave functions are expanded in plane-waves and we achieve an excellent scalability on high-performance computers. Being able to perform large-scale simulations involving several hundreds of electrons we systematically study the stopping of fast H and He atoms in bulk Al in order to address various scientific and computational issues. These include the influence of core electrons on the stopping and also how different charge states (e.g. He vs. He++) of the projectiles can play an important role. Prepared by LLNL under Contract DE-AC52-07NA27344.
5:30 AM - HH6.11
He and H Effect on Cavity Formation in Pure Iron and EB-welded F82H IEA Joint
Naoyuki Hashimoto 1 Tomoki Kubota 1 Tomonori Kimura 1 Somei Ohnuki 1 Shiro Jitsukawa 2
1Hokkaido University Sapporo Japan2Japan Atomic Energy Agency Tokai JapanShow Abstract
Reduced-activation ferritic/martensitic steel F82H has been developed as one of prime candidate materials for experimental fusion reactors. Some of the key issues are the effects of helium and hydrogen on the microstructure evolutions such as swelling, and on the mechanical properties such as fracture toughness or embrittlement. In this study, iron-base alloys and its electron-beam-welded joint have been irradiated by using an ion accelerator and a High Voltage Electron Microscope (HVEM) to examine synergistic effect of displacement damage and hydrogen or helium atoms on microstructure evolution, especially swelling behavior. Pre-injection of hydrogen (0~2000appm) and/or helium (0~1000appm) at RT, followed by electron irradiation in matrix at 400oC, were carried out for pure iron, F82H IEA and electron-beam-welded F82H IEA. Irradiation-induced cavities were observed in all the alloys. He-implanted alloys tended to exhibit higher number density and smaller mean size of the cavities compared to H-implanted alloys. In addition, in-situ electron irradiation experiment for H-implanted pure iron showed change in morphology and shrinkage of cavities. From these results, it is suggested that He effect on enhancement of cavity formation would be greater than H, while H could affect stability of cavities during irradiation. Multi-beam ion irradiation experiment with Fe+, He+, and H+ were also carried out at 400oC up to 10 dpa. Dual beam irradiation with Fe+ and H+ in pure iron resulted in lower number density and larger mean size of cavities compared to H-implanted and electron irradiation condition. Additionally, F82H and EB-welded F82H dual-beam -irradiated with Fe+ and H+ exhibited little formation of cavities. Therefore, it seemed that H effect on cavity formation in F82H would be much less than that in pure iron.
5:45 AM - HH6.12
Modeling Cavity Evolution in LWR Core Internal Components
Roger E. Stoller 1
1Oak Ridge National Laboratory Oak Ridge USAShow Abstract
In order to assess the potential for cavity swelling at end-of-life doses in LWR internal components fabricated from austentic stainless, it is necessary to develop a validated computational model that incorporates the relevant physical mechanisms. The basis for the current modeling activity is a comprehensive microstructural model that was developed to assess the swelling behavior of similar materials under fast neutron irradiation at elevated temperatures. The model employs the well-known reaction rate theory description of radiation damage formation and damage evolution, and accounts for the role of helium produced by nuclear transmutation in the nucleation and growth of voids. The model has been updated to account for recent advances in our understanding of primary damage production and point defect diffusion and reaction behavior obtained from atomistic simulations. Model validation has focused on the 250 to 325°C temperature range. The presentation will focus on the influence of critical irradiation and material parameters which would lead to the prediction of a level of swelling that is of engineering significance.
HH4: Nuclear Fuels II
Tuesday AM, November 27, 2012
Hynes, Level 1, Room 102
9:45 AM - *HH4.01
Nature and Behaviour of Point Defects in UO2 Based on an Experimental and Theoretical Study of Electrical and Atomic Transport Properties
Philippe Garcia 1 Elisabetta Pizzi 1 David Simamp;#233;one 2 Guido Baldinozzi 2 David Andersson 4 Jean-Paul Crocombette 2 Boris Dorado 3 Marjorie Bertolus 1 Michel Freyss 1 Guillaume Martin 1 Serge Maillard 1
1CEA/DEN Saint-Paul-Lez Durance France2CEA/DEN Gif-sur-Yvette France3CEA/DAM Arpajon France4LANL Los Alamos USAShow Abstract
Thermally or radiation induced transport properties impact practically all engineering aspects of nuclear oxide fuels, whether at the manufacturing stage, during in-reactor operation, or under long-term repository conditions. From a more fundamental standpoint, measuring transport properties is also a means of probing point or complex defects that are responsible for atomic migration. Although many studies relating to self-diffusion in UO2 have been carried out over the past forty years, these have not generally focussed on characterising these properties as a function of all the physical variables which determine it, i.e. temperature, the oxygen partial pressure and the impurity content, usually present in the form of bi- or tri-valent action impurities. In this talk, we show how electrical conductivity and self-diffusion property measurements may be combined in order to determine fundamental data relating to the nature of defects responsible for the property and their formation or migration energies. These data may then be compared to those obtained from first principles electronic structure calculations. The first part of the talk is dedicated to the development of a point defect model which captures the basic dependence of the electrical conductivity of UO2 upon temperature and oxygen partial pressure. The formation energies derived from this exercise are then compared to charged defect calculations using Density Functional Theory in the LDA+U approximation. The second part is concerned with oxygen self-diffusion and chemical diffusion coefficient measurements. We show that the point defect model is also compatible with the experimental data available. In the third part of the talk we examine uranium self-diffusion properties. The point defect model is specifically developed to account for uranium defects in a composition range close to stoichiometry. An analytical apparent activation energy is derived in this composition region and a numerical application is carried out based on basic formation and migration energy estimates obtained from first principles. The results compare favourably to existing data but highlight the need for additional experimental and theoretical work.
10:15 AM - HH4.02
Postirradiation Examination of Mixed Oxide Fuels for Actinide Transmutation
Heather J. MacLean Chichester 1 Douglas L. Porter 1 Steven L. Hayes 1
1Idaho National Laboratory Idaho Falls USAShow Abstract
Within the Fuel Cycle Research & Development program, six mixed oxide fuel compositions were fabricated and irradiated in the Advanced Test Reactor at the Idaho National Laboratory. Irradiation to 262 effective full power days resulted in fuel burnups ranging from approximately 5.8-8.4 at.% heavy metal. Fuel compositions included standard mixed oxide (MOX) fuel, (U0.80,Pu0.20)O2, and MOX fuel with minor actinides, (U0.75,Pu0.20,Am0.03,Np0.02)O2. These compositions are being studied in order to understand how fuel incorporating constituents from recycled light water reactor fuel would behave in fast reactors, with particular focus on transmuting actinides. Initial destructive postirradiation examinations of these compositions have been completed, including fission gas release, optical microscopy, microhardness testing, and burnup analysis. Initial calculations of fission gas release indicate that the fractional release of krypton and xenon fission gases ranges between 30% and 100% (complete release). Select postirradiation examination results, with a focus on fission gas release, will be presented and compared to results from the historical fuel performance database of fast reactor oxide fuels.
10:30 AM - HH4.03
Kinetic Monte Carlo Study of Interstitial Clusters in UO2+x
Rakesh Kumar Behera 1 Chaitanya S Deo 1 Taku Watanabe 2 Blas P Uberuaga 3 David A Andersson 3
1Georgia Institute of Technology Atlanta USA2Georgia Institute of Technology Atlanta USA3Los Alamos National Laboratory Los Alamos USAShow Abstract
Oxygen interstitials in UO2+x significantly affect thermophysical properties of the oxide nuclear fuel. Based on our previous work, we have analyzed the effect of mono-, di-, and other larger interstitial clusters on the oxygen ion diffusivity. First principles calculations are used to estimate the stability and energetic of different interstitial clusters. The estimated energies (migration, dissociation, etc.) with first principles are used to estimate the oxygen ion diffusivity in the kinetic Monte Carlo. We will discuss the effect of temperature and stoichiometry on the overall diffusivity of UO2+x. The computed diffusivities are compared with available experimental data. In addition we will present the sensitivity analysis of the associated energies on the diffusivity of oxygen interstitials. This work is funded by the DOE Office of Nuclear Energy&’s Nuclear Energy University Programs.
10:45 AM - HH4.04
Uranium Dioxide Films with Embedded Xenon
Igor Usov 1 Robert Dickerson 1 Patricia Dickerson 1 Marilyn Hawley 1 Darrin Byler 1 Kenneth McClellan 1
1Los Alamos National Laboratory Los Alamos USAShow Abstract
Xenon (Xe) is one of the major gas elements produced during nuclear fuel burning. It has long been known that Xe accumulation in the fuel pellet and Xe release to the plenum are detrimental to the fuel performance and safety and therefore must be well understood and controlled. Experimental data concerning description of Xe diffusion in UO2 -based nuclear fuels have a wide range of disparity such that it is difficult to verify modeling results for development of predictive nuclear fuel performance codes. Obtaining conventional UO2 samples containing even a few percent Xe accumulated in-pile is not possible without years of irradiation in a test reactor. An equally challenging problem, which relates to obtaining quantitative parameters governing Xe migration, is synthesis of reference UO2 samples with a uniform and controllable Xe concentration, separated from the various and complex other effects attendant to the fission process. The goal of this work was to fabricate and characterize such reference samples to isolate and quantify the inherent transport properties of Xe in UO2. We utilized ion beam assisted deposition (IBAD) to fabricate depleted UO2 (DUO2) films with embedded Xe atoms. The films were annealed at 1000 oC to induce Xe atom redistribution, and characterized before and after the annealing. Microstructural and chemical composition changes were examined by transmission electron microscopy (TEM), energy dispersive X-ray spectroscopy (EDXS), atomic force microscopy (AFM) and Rutherford backscattering spectroscopy (RBS). A detailed study of the deposition condition&’s influence on Xe content and morphology as well as the DUO2 film microstructure will be presented. Preliminary results detailing Xe-filled bubble growth and Xe atom diffusion in the DUO2 films will be presented.
HH5: Radiation Effects II - Ionic and Covalent Systems
Tuesday AM, November 27, 2012
Hynes, Level 1, Room 102
11:30 AM - *HH5.01
Effects of Ionization on Irradiation Damage Evolution and Recovery
William J Weber 1 2 Marie Backman 1 3 Yanwen Zhang 2 1 Flyura Djurabekova 3 Kai Nordlund 3 Marcel Toulemonde 4 Aurelien Debelle 5
1University of Tennessee Knoxville USA2Oak Ridge National Laboratory Oak Ridge USA3University of Helsinki Helsinki Finland4University of Caen Caen France5University Paris Sud Orsay FranceShow Abstract
The interaction of ions with solids results in energy loss to both atomic nuclei and electrons. At low energies, nuclear energy loss dominates, and irradiation damage occurs primarily by ballistic collisions. At high energies for fission products and swift heavy ions, electronic energy loss dominates, and the intense ionization can lead to latent track formation or recovery of existing irradiation damage. At intermediate ion energies, including energies of primary knock-on created by fast and fusion neutrons, ballistic and ionization energy losses are of similar magnitude and can lead to synergistic or competitive processes that that affect the evolution of irradiation damage. We have integrated experimental and computational approaches to investigate the separate and combined effects of nuclear and electronic energy loss on damage formation and recovery in several materials. Experimentally, we have shown that that there is a synergy between the nuclear and electronic energy loss on damage evolution in amorphous SiO2 at intermediate ion energies. Large scale molecular dynamics simulations, which include ballistic collisions and local heating based on the inelastic thermal spike model, have been employed to investigate the separate and combined effects of nuclear and electronic energy loss on damage production. These simulations demonstrate conclusively the additive effect on nuclear and electronic energy loss on damage production. On the other hand, ionization in Ca2La8(SiO4)6O2 from intermediate energy ion irradiation leads to competitive damage recovery processes that decrease damage production. In SiC, irradiation with intermediate energy ions leads to defect formation and amorphization; however, it has been shown that swift heavy ions can induce some recovery of such irradiation damage. Large scale molecular dynamics simulations confirm that swift heavy ions induce defect recovery and recrystallization in SiC that are well described by an inelastic thermal spike phenomenon.
12:00 PM - HH5.02
Size Dependence of the Radiation Induced Amorphization and Recrystallization Processes of Nanostructured CePO4 Monazite
Fengyuan Lu 1 Yiqiang Shen 2 1 Zhili Dong 2 Rodney Ewing 3 Jie Lian 1
1Rensselaer Polytechnic Institute Troy USA2Nanyang Technological University Singapore Singapore3University of Michigan Ann Arbor USAShow Abstract
The CePO4 monazite, with Ce as a surrogate for Pu, is considered as an important candidate to incorporate actinides for potential nuclear waste form applications. In this study, we synthesized nanostructured CePO4 monazite with different grain sizes ranging from 20 nm to over 100 nm. Separate and simultaneous displacive (1 MeV Kr2+) and ionizing (200 keV electrons) irradiations were conducted on the different sized CePO4 in order to investigate their radiation response. In situ TEM observation revealed that CePO4 nanoparticles can be amorphized by 1 MeV Kr2+ irradiation with better radiation tolerance than their bulk counterpart, whereas 200 keV electron-beam irradiations can induce recrystallization of the pre-amorphized CePO4. An abnormal behavior with a strong size effects on the radiation response of nano-sized CePO4 was identified. The 20 nm sized CePO4 exhibits lower radiation tolerance than the larger 40 nm sized ones, probably due to excess surface energy of reduced sized CePO4 which may alter the energy difference between the amorphous and crystalline phases and consequently affect the radiation stability. A higher recrystallization rate was observed in smaller sized CePO4 under ionizing irradiations, which may result from the higher specific surface area that can provide preferential recrystallization sites for CePO4 grains. When irradiated with simultaneous ionizing electrons and displacive ions, CePO4 displayed high tolerance against amorphization, implying a desired radiation performance for its nuclear waste form application. The remarkable size dependence of the displacive radiation-induced amoprhization and ionizing radiation-enhanced recrystallization processes for nanostructured CePO4 indicates the existence of a critical grain size for optimized radiation tolerance.
12:15 PM - *HH5.03
Transmission Electron Microscope Study of Defect Formation and Accumulation in Ceramic Oxides Irradiated with Swift Heavy Ions
Syo Matsumura 1 Tomokazu Yamamoto 1 Kazuhiro Yasuda 1 Seiya Takaki 1
1Kyushu University Fukuoka JapanShow Abstract
Radiation-induced defect formation and accumulation due to high-density electronic excitation is one of the essential issues to understand the radiation damage processes induced by fission fragments. In the present talk, we will give an overview of our recent transmission electron microscope (TEM) studies on defect formation and accumulation in ceramic oxides irradiated with swift heavy ions. It is well known that high-density electronic excitation induced by irradiation with swift heavy ions results in formation of columnar defects called as ion tracks in oxide ceramics. High resolution TEM observation of the ion tracks formed in MgAl2O4 has shown that the crystalline lattice survives irradiation with 340 MeV Au ions even in the central core regions along the ion trajectories. Here the electronic stopping power was evaluated to be 34 keV/nm. The TEM contrast of ion tracks appearing in conventional bright-field (BF) images depends on excitation of Bragg reflections. Dark columnar contrast clearly appears along the ion tracks in an inclined view with g=220, whereas it becomes faint when the Bragg position is on g=440. The contrast variation suggests structural deviation from the spinel-type crystal structure inside the ion tracks. The strain field contrast stands out under the latter condition, indicating the formation of extended defect aggregations in the matrix. TEM BF-imaging with defocus and HAADF-STEM have shown reduction of atomic density, namely formation of a high concentration of vacancies at the cores of ion tracks. The number density of ion tracks increases proportionally with ion fluence in the earlier stage of accumulation but it saturates to keep a constant after reaching 1016 ions/m2. This result is discussed in terms of a balance between the formation and the annealing of the ion tracks under irradiation. The annealing range to induce the recovery is evaluated to be 7-9 nm in radius. Both electron channeling X-ray analysis (HARECXS) and electron diffraction have revealed that cations tend to evacuate from the tetrahedral sites and reside preferentially on the octahedral sites after significant overlaps of radiation damages. The results on CeO2 with the fluorite structure also will be referred in the talk.
12:45 PM - HH5.04
Simulating Radiation-induced Defect Formation in Pyrochlores
David Gunn 1 John Purton 1 Ilian Todorov 1
1Science amp; Technology Facilities Council Daresbury United KingdomShow Abstract
The disposal and safe storage of nuclear waste is a significant challenge for the global community. Several of the radionuclides generated through the nuclear fuel cycle, such as 239Pu and 235U, have long half lives (24,100 years and 7x108 years respectively) and careful choice of suitable immobilisation matrices is crucial to prevent any environmental contamination. Such an immobilisation material must be able to withstand prolonged heavy ion particle bombardment while maintaining structural integrity. Pyrochlore-type compounds have been proposed as suitable host matrices for this purpose, and great attention has been paid to members of the series Gd2(ZrxTi2-x)O7 (0le;xle;2). The radiation tolerance of this series increases with increasing zirconium content, and the healing process in the zirconate is expected to be faster than for the titanate as it does not undergo an amorphous transition upon radiation damage and is a fast ion conductor. We propose a new set of Buckingham potentials, specifically tailored for looking at this Gd2(ZrxTi2-x)O7 series. The accuracy and robustness of these new potentials is demonstrated by calculating and comparing values for a selection of point defects with those calculated using a selection of other published potentials and our own ab initio values. Frenkel pair defect formation energies are substantially lowered in the presence of a small amount of local cation disorder. The activation energy for oxygen vacancy migration between adjacent O48f sites is calculated for Ti and Zr pyrochlores using an improved tangent nudged elastic band method. This energy is lower for the non-defective Ti than for the Zr pyrochlore by ~0.1 eV, consistent with the majority of the potentials tested. The effect of local cation disorder on the VO48f → VO48f migration energy is minimal for Gd2Ti2O7, while in contrast the migration energy is lowered typically by ~43% for Gd2Zr2O7. Since the healing mechanisms of these pyrochlores are likely to rely upon the availability of oxygen vacancies, the healing of a defective Zr pyrochlore is predicted to be faster than for the equivalent Ti pyrochlore.
Gianguido Baldinozzi, CNRS-Ecole Centrale Paris
Maria Jose Caturla, Universidad de Alicante
Chaitanya S. "Deo", Georgia Institute of Technology
Chu-Chun Fu, "CEA-Saclay DEN/DMN/SRMP"
Kazuhiro Yasuda, Kyushu University
Yanwen Zhang, Oak Ridge National Laboratory
HH9: Radiation Effects III - Insulators
Wednesday PM, November 28, 2012
Hynes, Level 1, Room 102
2:30 AM - *HH9.01
Damage and Recovery in Ceramics - Is it Predictable?
Karl R Whittle 1 2
1University of Sheffield Sheffield United Kingdom2Australian Nuclear Science and Technology Organisation Sydney AustraliaShow Abstract
Ceramics have been shown to have vastly different responses to radiation damage, in some cases recovery can be rapid, in others it can be slow. What are the likely contributors to the recovery rate is the subject of much research, both experimental and simulation in nature. Using model systems it has been possible to examine the contributions of both bonding/chemical composition and structure. Primarily this through the modification of either composition or structure, i.e. modifying one with minimal, or zero, change in the other. To fully appreciate the effects of radiation damage on a system, the undamaged, or equilibrium, structure itself must be understood. The information derived from such studies can be integrated into understanding further recovery mechanisms from damage. For example, why are certain structures adopted for a given composition, when others are possible, will impact on the recovery, and more importantly, the rate from damage. For example in some systems the recovery from damage has been found to be non-linear with changes in composition, whereas in others the reverse is true. In such systems it is found that the recovery from damage initially increases with a change in composition, whereas with continued change the recovery rate begins to decrease. What are the competing drivers within the system that modifies how a system responds to damage? This work uses two examples to outline how recovery from damage can be predicted by examining structures/bonding of materials prior to irradiation. It will also show one in which the process of predicting damage recovery is not only complex, but in some cases unexpected.
3:00 AM - HH9.02
Radiation Damages on the Porous Properties of Mesoporous Silica Thin Films and Bulk Materials
Sandrine Dourdain 1 Xavier Deschanels 1 Guillaume Toquer 1 Stamp;#233;phane Pellet-Rostaing 1 Agnes Grandjean 1
1CEA Marcoule 30207 Bagnols sur Ceze FranceShow Abstract
Mesoporous silicas are highly potential materials for applications in the nuclear field for separation, recycling or confinement of nuclear wastes. They present a porous network that may be organized (SBA, MCM) or disordered (Vycor glass). For these intended applications it is essential to know the behavior of these materials under irradiation, as it might induce modifications of the porous properties as for example closure or destruction of the micro- or meso-porosity. For this purpose different ion irradiation experiments, with different regimes of energy deposition (electronic, ballistic), were implemented on these mesostructured materials, playing with ions energy, fluence and nature (Xe, Au, He, C, Ar,hellip;). The specific case of mesoporous thin films was investigated with the adapted characterization techniques as X ray Reflectivity and cross sectional Scanning Electron Microscopy, and compared to radiation effects on bulk mesoporous materials as Vycor glass (analysed with X rays and Nitrogen adsorption measurements), to access complementary informations on surface and volume effects. After an overview of the preliminary results and taking into account the irradiation effects described in the litterature for dense and porous silica materials and the available models, the presentation will attempt to answer the question of the existence of a relationship between the rigidity of the silica network and its mesoporous structure, with the radiation damages that are induced.
3:15 AM - HH9.03
Spectroscopic Investigation of Ion Beam Irradiation Induced Structural Modifications in Model Nuclear Waste Glasses
Amy Sarah Gandy 1 Martin Christopher Stennett 1 Neil Christian Hyatt 1
1University of Sheffield Sheffield United KingdomShow Abstract
In the UK, alkali borosilicate glasses are used to vitrify high level waste (HLW) produced by reprocessing of spent nuclear fuel. HLW contains fission products and minor actinides which continue to undergo radioactive decay in the wasteform for up to 1 million years. Other cations such as Fe and Zr are also present in the waste stream and require incorporation into the final wasteform. Actinides undergo α-decay events with the formation of α-particles (He nuclei) and energetic (~100KeV) daughter recoil nuclei. Energy is transferred from the energetic recoil nuclei to other atoms in the glass via elastic interactions, resulting in atomic displacements which form collision cascades. Accumulation of this ballistic damage can lead to migration of alkali ions, resulting in changes in glass network polymerisation, and changes in cation valance state. These changes can affect wasteform durability and since the wasteform acts as the final barrier against radionuclide release into the environment, it is important that the effects of α-decay on the structure of the glass are understood. In this study, heavy ion implantation (e.g. 450KeV Kr irradiation) was used to provide an analogue for the α-recoil damage and samples were irradiated at room temperature with a dose relevant to vitrified product lifetimes (0.1 - 1.0 displacements per atom (dpa)). The effects of simulated α-recoil damage were investigated by probing the speciation and valence of Fe as a network intermediate, in analogue nuclear waste glasses, as a function of glass composition. In this contribution we report on the effect of heavy ion implantation on the Fe oxidation states and co-ordination environment, in various model glasses, using X-ray absorption and Raman spectroscopies.
3:30 AM - HH9.04
The Impact of Thermal Activation on Defect Production in Rutile
Marc Robinson 1 Nigel A Marks 1 Greg R Lumpkin 2
1Curtin University Perth Australia2Australian Nuclear Science and Technology Organisation Sydney AustraliaShow Abstract
In the development of current and future nuclear materials, it is key to determine the functionality of a material in the environment of its intended use. Central to this is principle is the requirement to understand how a material's irradiation response is affected by temperature. Using computer simulation it is possible to capture the fundamental atomic-scale processes responsible for the accumulation of damage within a material. In this work, we systematically study the initial phase of radiation events to determine the threshold displacement energy (Ed) using extensive molecular dynamics simulations. To quantify the effect of temperature on defect production, the simulations involve rigorous sampling of the impact energy and direction of Primary Knock-on Atoms (PKAs) at a range of temperatures from 50 to 1200 K. In application to rutile TiO2, defect production on the oxygen sublattice is found to be significantly affected by increases in temperature. This leads to a marked reduction in the number of residual defects from both PKA species across the energy range studied. In addition, a shift in the oxygen value of Ed is observed from 18 eV at 300 K to 53 eV at elevated temperatures. This is attributed to oxygen Frenkel pair recombination mechanisms that are thermally activated during the simulations. Significantly, transitions of this kind would occur well within the timescale of experimental techniques of determining Ed. These techniques report a higher value of Ed for oxygen indicating the short-lived defects present in the simulations are recombining before experimental detection takes place. This work highlights the importance of understanding the link between temperature and timescale when calculating values of Ed. If comparisons are to be made between experiment and simulation or if Ed is to be used in models of radiation damage, it is imperative to know the context in which the value of Ed was determined.
HH10/LL12: Joint Session: Radiation Effects
Wednesday PM, November 28, 2012
Hynes, Level 1, Room 102
4:15 AM - HH10.01/LL12.01
Novel Fast Reactor Fuels Manufactured by Freeze Casting
William J. Goodrum 1 Philipp M. Hunger 1 Shih-Feng Chou 1 Joan Burger 1 Amanda Lang 2 Thomas Gage 2 Clarissa Yablinsky 2 Todd R. Allen 2 Ulrike G.K. Wegst 1
1Dartmouth College Hanover USA2University of Wisconsin - Madison Madison USAShow Abstract
Advanced burner reactors are designed to reduce the amount of long-lived radioactive isotopes in spent nuclear fuel. The input feedstock for advanced fuel forms derives from either recycled light water reactor fuel, or recycled fast burner reactor fuel. In order to achieve higher performance and increase operational safety, these advanced reactors require novel fuel concepts, made from new materials. One promising pathway to improve fuel performance is the creation of metal or ceramic scaffolds, into which fuel may be placed with greater precision than in existing CERMET fuels. In this presentation, the design and manufacture of novel structures by “freeze casting” will be described. Freeze casting (or “ice templating”) is a directional solidification process ideal for the production of both metal and ceramic fuel scaffolds. This process inherently allows for the manufacture of a range of custom-tailored fuel pellet designs. The mechanical, thermal, and neutronic properties of both metal and ceramic scaffolds will be compared and contrasted, in order to shed light on the performance and lifetime behavior of these novel fuel designs.
4:30 AM - HH10.02/LL12.02
Ion Beam Irradiation Effects in NZP-structure Type Ceramics
Daniel J Gregg 1 Inna Karatchevtseva 1 Joel Davis 1 Michael James 3 Gordon I. Thorogood 1 Pranesh Dayal 1 Benjamin Bell 4 Matthew Jackson 4 Mihail Ionescu 2 Gerry Triani 1 Ken T. Short 1 Gregory R. Lumpkin 1 Eric R. Vance 1
1ANSTO Kirrawee DC Australia2ANSTO Kirrawee DC Australia3ANSTO Kirrawee DC Australia4Imperial College London London United KingdomShow Abstract
Sodium zirconium phosphate (NZP) type ceramics accommodate approximately 42 elements of the periodic table including most fission products derived from nuclear power plant fuel. As such, NZP-structure type ceramics have considerable potential as host materials for the immobilization of radioactive waste as well as candidate inert matrices for minor actinide burning. It is therefore important to investigate the behaviour of this material under irradiation conditions in order to verify its long-term stability. In this study strontium zirconium phosphate (an NZP-type structure ceramic) has been irradiated with gold and helium ions to simulate the consequences of alpha decay. The effects of the irradiation on the structural as well as macroscopic properties (e.g. density and hardness) are investigated using grazing-incidence X-ray diffractometry, Raman spectroscopy, scanning electron and atomic force microscopy, and nano-indentation. Irradiation by gold ions results in significant changes to the crystalline structure and hardness. After a fluence of 1015 gold ions/cm2, strontium zirconium phosphate undergoes structural amorphization, a volume reduction, and an increase in hardness. These results as well as the results from He-ion irradiation are discussed with regard to the application of NZP-structure type ceramics as inert matrices for minor actinide burning or as host materials for the immobilization of radioactive waste.
4:45 AM - HH10.03/LL12.03
Ion Beam Irradiation of Crystalline ABO4 Compounds
Massey de los Reyes 1 Daniel Gregg 1 Robert Elliman 2 Nestor Zaluzec 3 Robert Aughterson 1 Gregory Lumpkin 1
1ANSTO Sydney Australia2ANU Canberra Australia3ANL Chicago USAShow Abstract
Fergusonite and scheelite-structured ABO4 ternary oxides are an important class of materials owing to their technological applicability and geological significance. In spite of their growing interest as potential wasteform ceramics, only very little is known about their behaviour under irradiation in regards to other ABO4 analogues such as zircon and monazite. To this purpose, we have studied and compared the effects of ion-beam irradiation on compounds LaVO4, YNbO4 and CaWO4 by 1 MeV Kr+ ions as a function of irradiation temperature (50 - 600K). Resulting critical temperatures for amorphisation (Tc) differ slightly for LaVO4 and YNbO4 each with a Tc of 400K and 450K respectively. CaWO4 shows stronger amorphisation ‘resistance&’ and has a Tc of 200K. The susceptablity toward amorphisation and disorder in each structure is discussed in terms of their structural parameters as well as the stopping powers, displacement energies, and defect energies of the materials. The phase transitions that occur between tetragonal scheelite and monoclinic fergusonite will also be highlighted.
5:00 AM - HH10.04/LL12.04
Understanding the Metamict State in Titanate Ceramics for Nuclear Waste Immobilisation Using Molecular Dynamics and Connectivity Topology Analysis
Henry R Foxhall 1 Karl P Travis 1 John Harding 1 Scott L Owens 2 Linn W Hobbs 3 4
1University of Sheffield Sheffield United Kingdom2National Nuclear Laboratory Risley United Kingdom3Massachusetts Institute of Technology Cambridge USA4Massachusetts Institute of Technology Cambridge USAShow Abstract
This study presents structural analysis of crystalline and radiation-damaged zirconolite, CaZrTi2O7, and pyrochlore, Gd2Ti2O7, both potential actinide-accommodating nuclear waste materials, using molecular dynamics (MD) and connectivity topology analysis - a powerful method for describing both crystalline structures and their metamict or amorphous analogues, because it places no reliance on symmetry operators or periodic translation, both of which vanish upon introduction of disorder to a material. The work establishes characteristic topological differences in the connectivity of each structure and finds evidence that amorphization induced by alpha-recoil displacement cascades still retains certain short- and intermediate-range ordered configurations, particularly for Ti atoms. [TiOx] polyhedral edge-sharing chains are observed in the metamict state in both materials, which may act to stabilize the radiation-damaged structure and prevent recovery of the initial crystalline phase. We also present an assessment of the predicted amorphizability of zirconolite based on the topological constraints imposed by its structure, finding that the varying structural rigidity of the layers in the structure is crucial to its amorphizability potential. The hexagonal tungsten bronze structure [TiOx] layer in particular provides weak constraints that are responsible for zirconolite&’s comparative ease of amorphization.
5:15 AM - HH10.05/LL12.05
The Effect of Pressure on the Radiation Tolerance of the Polymorphs of TiO2
Meng J Qin 1 Simon Charles Middleburgh 1 Eugenia Kuo 1 Karl R Whittle 1 Nigel A Marks 2 Marc Robinson 2 1 Greg R Lumpkin 1
1ANSTO Lucas Heights Australia2Curtin University of Technology Perth AustraliaShow Abstract
Molecular dynamics simulations using thermal spikes have been carried out to investigate the effect of pressure on the time-dependent generation of defects under irradiation in the three common polymorphs of TiO2: rutile, anatase and brookite. The effect of crystal structure on the tollerance to radiation damage was first investigated, highlighting the experimental observation that the rutile phase is the most tollerant to damage. The density of the phases was then varied and the same thermal spike methodology repeated with some interesting results suggesting a strong correlation between density and radiation tollerance.
5:30 AM - HH10.06/LL12.06
Advanced Measurement Techniques for Irradiated Nuclear Fuels and Materials
John Rory Kennedy 1
1Idaho National Laboratory Idaho Falls USAShow Abstract
In the realm of radioactive nuclear materials, a major challenge to the development of materials is the measurement of the properties for which the material is being developed. For example, the phenomenon of microstructure evolution of a nuclear fuel in reactor is well known but the details of the effects of the change on the behavior of such important issues as thermal conductivity, mechanical properties, and phase formation have not been quantified at the grain size level. There is a strong need to develop or adapt advanced instrumentation for measurements on radioactive materials. Idaho National Laboratory has an ongoing effort to develop or adapt a variety of measurement techniques to highly radioactive materials. A laser based device termed the Scanning Thermal Diffusivity Microscope, conceived and developed over the past few years, has recently been installed in a hot cell where examinations of fresh and irradiated fuel samples have begun in order to profile the thermal diffusivity of fuels and materials at 50µm spatial resolution. A second generation instrument close to implementation will soon give thermal conductivity values at 5-10 µm. The unique application of dual-beam focused ion beam (FIB) to the preparation of highly radioactive material samples has become an exceedingly useful tool for determining 3D grain orientation (EBSD), mechanical properties by nano/micro indentation or compression testing, microstructure through transmission electron microscopy, and nano-scale element distribution by atom probe tomography. This contribution will present the current state of the implementation plan of these instruments to highly radioactive fuels and materials and examples from ongoing irradiated fuels and materials studies will be given.
5:45 AM - HH10.07/LL12.07
Measuring Parameters of Dynamic Annealing in Ion-irradiated Solids
S. Charnvanichborikarn 1 M. T. Myers 1 2 L. Shao 2 Sergei O. Kucheyev 1
1Lawrence Livermore Nat'l Lab Livermore USA2Texas Aamp;M University College Station USAShow Abstract
Under ion irradiation, all crystalline materials display some degree of dynamic annealing when defects experience evolution after the thermalization of collision cascades. The exact time and length scales of such defect relaxation processes are, however, unknown even for Si at room temperature. Here, we propose a method to measure effective diffusion lengths and relaxation times of mobile defects that dominate the formation of stable post-irradiation disorder. A defect lifetime of about 5 ms and a characteristic defect diffusion length of about 30 nm are measured for Si at room temperature, essentially independent of the average density of ballistic collision cascades. Defect relaxation appears to be dominated by a second order kinetic process. We discuss implications of these findings for the development of predictive models of radiation damage buildup in solids. This work was performed under the auspices of the U.S. DOE by LLNL under Contract DE-AC52-07NA27344.
HH7: Plasma Facing Materials
Wednesday AM, November 28, 2012
Hynes, Level 1, Room 102
9:30 AM - *HH7.01
Plastic Localization in Irradiated Ferritic Systems: New Insights from Modeling and Simulation
Jaime Marian 1 Tom Arsenlis 1 Nathan Barton 1 Moon Rhee 1 Greg Hommes 1
1Lawrence Livermore Nat'l Lab Livermore USAShow Abstract
Low temperature irradiation of crystalline materials is known to result in hardening and loss of ductility, which limits the usefulness of candidate materials in harsh nuclear environments. In bcc metals, this mechanical property degradation is caused by the interaction of in-grown dislocations with irradiation defects, particularly small dislocation loops resulting from the microstructural evolution of displacement cascades. In this work, we present a multi scale model encompassing dislocation dynamics (DD) simulations, crystal plasticity, and finite element (FE) simulations of bcc Fe containing various concentrations of dislocation loops produced by irradiation in an attempt to gain insight into the processes that lead to hardening and embrittlement. The DD simulations reveal a transition from homogenous to highly localized deformation at a critical loop density. Above it, plastic flow proceeds heterogeneously, creating defect-free channels in its wake. These simulations are then used to calibrate a tensorial crystal plasticity model capable of reaching strains in excess of 10%. The calibrated crystal plasticity model is used as the constitutive relation in FE simulations of polycrystalline irradiated Fe systems.
10:00 AM - HH7.02
Capabilities of Nanostructured W as Plasma Facing Material in Future Fusion Reactors
R. Gonzalez-Arrabal 1 N. Gordillo 1 2 A. Rivera 1 I. Fernandez-Martinez 3 4 M. Panizo-Laiz 1 J. Y. Pastor 5 E. Tejado 5 K. Saravanan 6 F. Munnik 6 J. M. Perlado 1
1Polytechnic University of Madrid (UPM) Madrid Spain2CEI Campus Moncloa, UCM-UPM Madrid Spain3Polytechnic University of Madrid (UPM) Madrid Spain4Institute of Microelectronics of Madrid, IMM-CNM-CSIC Madrid Spain5Polytechnic University of Madrid (UPM) Madrid Spain6Forschungszentrum Dresden-Rossendorf Dresden GermanyShow Abstract
One of the challenges in the design of future nuclear power plant is to develop materials capable to resist in the hostile environment of a fusion reactor. Because of its low sputtering yield, low-activation, high melting point, high thermal conductivity and low thermal expansion, tungsten is one of the most attractive materials proposed for first wall applications in nuclear fusion reactors [1-3]. Even when W is assumed to be the best candidate as plasma facing material (PFM), some limitations have been identified that have to be defeated in order to fulfil specifications i.e. an important point of concern to the light species behavior (H, D, T and He). Nowadays some strategies to overcome these limitations are being investigated . In this work we focus on the study of the capabilities of nanoW as PFM. Firstly, we report about DC magnetron sputtering deposition procedure, presenting the dependence of sample microstructure on deposition parameters. Microstructural characterization studies by XRD,TEM and SEM evidence that nanostructured samples are polycrystalline and are composed of columns with a diameterbetween 50 and 200 nm. Then, the thermal properties (conductivity and stability) are studied. These results illustrate that the column diameter does not significantly increases in the temperature range up to 400 C. For temperatures higher than 800 C the adhesion between W and the steel substrate has identified to be a major problem. Finally, the light species behaviour is characterized as a function of sample microstructure and implantation conditions. The role of the synergetic effects,when the samples are simultaneously exposed to different particle irradiation, in the light species behaviour is addressed. For this purpose resonant nuclear reaction (RNRA) experiments were performed using the H(15N,α)12C nuclear reaction in nanostrcutred (nW) and polycrystalline (pW) samples implanted with (i) H at an energy of 170 keV, (ii) sequentially implanted with C at an energy of 665 keV and H at 170 keV and (iii) simultaneously implanted with C and H at the above described energies. Implantations were carried out at a fluence of 5e16at/cm2 and at two different temperatures RT and 400 C. RNRA data evidence that the highest H retention is observed for the C and H co-implanted samples, being the lower one measured for those samples implanted only with H. In general, the H retention is higher for nW than for pW samples. Moreover, increasing the irradiation temperature up to 400 C drives the H to completely out diffusion in