Symposium Organizers
Jarir Aktaa Forschungszentrum Karlsruhe GmbH
Maria Samaras Paul Scherrer Institute
Magdalena Serrano de Caro Lawrence Livermore National Laboratory
Maximo Victoria Polytechnic University of Madrid
Brian Wirth University of California-Berkeley
Tuesday PM, November 28, 2006
Berkeley (Sheraton)
9:30 AM - **JJ1.1
Atomic Scale Modelling of the Primary Damage State of Irradiated UO2 Matrix.
Laurent Van Brutzel 1 , Jean-Paul Cocombette 2
1 DTCD/SECM/LCLT, CEA-Marcoule, Bagnols sur cèze France, 2 DMN/SRMP, Cea-Saclay, Gif sur Yvette France
Show AbstractLarge scale classical molecular dynamics simulations have been carried out to study the primary damage state due to alpha-decay self irradiation in UO2 matrix. Simulations of energetic displacement cascades up to the realistic energy of the recoil nucleus (80 keV) provide new informations on the defect production, their spatial distribution and their clustering. The discrepancy with the classical linear theory NRT (Norton-Robinson-Torrens) law on the creation of the number of point defects is discussed. A new empirical relationship between Frenkel pair production and damage energy is proposed. Study of cascade overlap sequence shows a saturation on the number of point defects created as the dose increases. Toward the end of the overlap sequence, large stable clusters of vacancies are observed. A athermal diffusion coefficient is also estimated from these results.
10:00 AM - JJ1.2
Reduction of Radiation Damage Near Cu-Nb Interfaces in Cascade Simulations.
Michael Demkowicz 1 , Richard Hoagland 1 , Amit Misra 2 , Yun-Che Wang 2
1 MST-SPR: Structure-Property Relations, Los Alamos National Laboratory, Los Alamos, New Mexico, United States, 2 MST-CIN: Center for Integrated Nanotechnologies, Los Alamos National Laboratory, Los Alamos, New Mexico, United States
Show Abstract10:15 AM - JJ1.3
Parallel Kinetic Monte Carlo.
Enrique Martinez 1 2 , Jaime Marian 2 , Malvin Kalos 2
1 DENIM, UPM, Madrid, Madrid, Spain, 2 CMS, LLNL, Livermore, California, United States
Show AbstractThe parallelization of kinetic Monte Carlo (kMC) codes has always been a major bottleneck in modelling cascade evolution and damage accumulation due to the inherent spatial and temporal heterogeneities found in this type of calculations. Previous attempts to parallelize kMC have always relied on spatial decompositions whereby each subdomain is evolved separately in time. The problem with evolving each spatial domain independently from each other is that a time asynchronicity develops due to a mismatch in the total rate length, Ri, assigned to each subdomain i. This is an issue that has proven particularly difficult to solve even though some clever solutions have been proposed. In this presentation we will propose a method based on a constrained spatial decomposition that minimizes the ratios between domain rates. The deviation from unity of these rations arises from a non-uniformity of the particle distribution density in the system. To ensure time synchronicity we extend each domain rate up to a prescribed value, Rq>Ri, which is the same for all spatial subdomains. This extension is performed by resorting to ‘dummy’ events whose effect is nil in the kinetics of the system. In this way, the timestep τ is sampled from the same distribution exp(-Rq.τ) for all spatial subdomains such that the time gain with respect to the serial treatment is of the order of ΣRi/Rq. We will demonstrate that our method provides an exact solution of the master equation with and without volumetric sinks and examples of scalability and error with respect to analytical solutions will be shown for the diffusion equation under different boundary conditions.
JJ2/GG7: Joint Session: Radiation Damage
Session Chairs
Maria Caturla
Sergei Dudarev
Tuesday PM, November 28, 2006
Constitution B (Sheraton)
11:00 AM - JJ2.1/GG7.1
Modeling Radiation Damage Evolution in Fe-Cr Alloys.
Brian Wirth 1 , Hyon-Jee Lee 1 , Jae-Hyoek Shim 2 , Kevin Wong 1
1 Nuclear Engineering, University of California, Berkeley, Berkeley, California, United States, 2 , Korea Institute of Science and Technology, Seoul Korea (the Republic of)
Show Abstract11:15 AM - JJ2.2/GG7.2
Multiscale Approach for Understanding Advanced Materials.
Wolfgang Hoffelner 1 , Maria Samaras 1 , Manuel Pouchon 1 , Jia Chao Chen 1 , Maximo Victoria 1 , Annick Froideval 1 , Botond Bako 1 , Roberto Iglesias 1
1 , Paul Scherrer Institiute, Villigen Switzerland
Show Abstract11:30 AM - JJ2.3/GG7.3
From Interstitial Clusters to Interstitial Loops in Iron: A Multiscale Approach Based on First Principles.
Mihai-Cosmin Marinica 1 , Francois Willaime 1
1 SRMP, CEA/Saclay, Gif-sur-Yvette France
Show AbstractInterstitial-type defects formed by the clustering of self-interstitials produced under irradiation have rather peculiar properties in alpha-iron. Very small interstitial clusters or loops are formed of <110> dumbbells, whereas larger clusters have either a <111> or a <100> orientation. This contrasts with other BCC metals where they are predominantly <111>. The competition between these different orientations raises the question of their relative stabilities as function of size and temperature, and of the transformation mechanism from <111> to <100> orientations observed experimentally. Their mobilities are also a key issue in the interpretation of experiments. We have addressed these questions by combining first principles and empirical potential approaches. We have determined the stability and mobility of small self-interstitial defects by ab-initio calculations performed on cells containing up to 250 atoms using the SIESTA code. We then used these results to fit and validate a new semi-empirical potential for iron. This potential allows to access dynamical properties and larger sizes. (i) For small clusters, we performed a systematic search of possible configurations, including non-parallel ones, and of migration/rotation mechanisms. (ii) Using lattice-dynamics we studied the vibrational properties of small clusters: low frequency modes have been evidenced in <111>-type defects; the associated large vibrational entropy is shown to have a significant effect on the relative stabilities at finite temperature. (iii) In connection with experiments we have investigated the properties of loops with up to 1000 defects: the relative stabilities as function of Burgers vector, their migration energies, and the loop-loop and loop-surface interactions. We discuss the dependence on the potential of these results.
11:45 AM - JJ2.4/GG7.4
Role of Trapping Impurities on He Desorption and Clustering in Irradiated a-Fe.
Christophe Ortiz 1 , María José Caturla 1 , François Willaime 2 , Chu Chun Fu 2
1 Departamento de Física Aplicada, University of Alicante, E-03690 Alicante, Alicante, Spain, 2 Service de Recherches Métallurgiques, CEA/Saclay, F-91191 Gif-sur-Yvette France
Show AbstractIt is well-known that impurities affect the migration of intrinsic point defects in metals. For instance, carbon is a common impurity in Fe that significantly retards diffusion of vacancies. Under fusion irradiation conditions, high levels of He are produced by transmutation reactions. This element strongly interacts with vacancies produced during irradiation and agglomerate into stable He-vacancy clusters that can deteriorate the mechanical properties of the material. A physically-based model accounting for the interactions between He, point defects (interstitials and vacancies) and trapping impurities is therefore necessary to understand and predict damage evolution in Fe.We have used a multi-scale approach to predict the evolution of He in the presence of impurities in irradiated Fe. Density Functional Theory (DFT) calculations were performed to investigate the migration mechanisms and to determine the activation energies of the different atomistic processes. The influence of impurities - such as carbon - on the binding energies of small He-vacancy clusters was also studied. Using the information obtained by DFT a physically-based model was developed and implemented in a kinetic Monte Carlo (kMC) code to follow the evolution of He in Fe. In addition, a model based on the rate theory (RT) was developed in order to achieve larger simulation times and volumes. Results obtained with this model, which is based on a mean field approximation are compared to those obtained with kMC. Using this multiscale approach, the simulation results are used to interpret the different stages of thermal He desorption experiments and to determine the predominant migration mechanism. The influence of impurities which affect the diffusion of point defects or modify the binding energies of He-vacancy type clusters is also studied.
12:00 PM - **JJ2.5/GG7.5
Multiscale Modelling of bcc-Fe Based Alloys for Nuclear Applications.
Lorenzo Malerba 1
1 RMO, SCK-CEN, Mol Belgium
Show AbstractUnderstanding the basic mechanisms that determine microstructure changesin neutron irradiated steels is vital for a safe lifetime management of existing nuclear reactors and a safe design of future nuclear options. Low-alloyed ferritic steels containing Cu, Ni, Mn and Si as principal solute atoms are used as structural materials for current reactor vessels, while high-Cr ferritic-martensitic steels will be used in future nuclear options. The microstructural evolution under irradiation in alloys is decided by the interplay between defect formation and thermodynamic driving forces, together determining the appearance of phase transformations (precipitation, segregation, ...) and favouring or delaying the nucleation and growth of point-defect clusters, their diffusion and their mutual recombination or removal at sinks. A reliable description of the production, evolution and accumulation of radiation damage must therefore start from the atomic level and requires being able to describe multicomponent systems for timescales ranging from few picoseconds to years. This goal demands firstly the fabrication of interatomic potentials for alloys that must be both consistent with the thermodynamic properties of the system and capable of reproducing correctly the characteristic solute-point defect interactions, versus ab initio or experimental data. Secondly the performance of extensive molecular dynamics (MD) simulations, to grasp the main mechanisms of defect production, diffusion, mutual interaction, and interaction with solute atoms and impurities. Thirdly, the development of simulation tools capable of describing the microstructure evolution beyond the timeframe and lengthscale of MD, while reproducing as much as possible the atomic-level origin of the mechanisms governing the evolution of the system, including phase changes.In this presentation the results of recent efforts made in this direction in the case of Fe-Cu and Fe-Cr alloys, as basic model alloys for the description of steels of technological relevance, are highlighted. In particular, advanced techniques to fit interatomic potentials consistent with thermodynamics are proposed and the results of their application to the mentioned alloys are presented. The results of the use of advanced potentials to study the effect of high concentrations of solute atoms on self-interstitial cluster mobility and their correlation to changes in macroscopic properties, such as swelling, are summarised. And the development of advanced methods, based on the use of artificial intelligence, to improve both the physical reliability and the computational efficiency of kinetic Monte Carlo codes for the study of point-defect clustering and phase changes beyond the scale of MD, is reported. These recent progresses bear the promise of being able, in the near future, of producing reliable tools for the description of the microstructure evolution of realistic model alloys under irradiation.
12:30 PM - JJ2.6/GG7.6
Ab Initio and Kinetic Rate Theory Modeling of 316SS with Oversized Solute Additions on Radiation-Induced Segregation.
Micah Hackett 1 , Gary Was 1
1 Nuclear Engineering and Radiological Sciences, University of Michigan, Ann Arbor, Michigan, United States
Show AbstractDeleterious effects of radiation in nuclear reactor systems cause material degradation and the potential for component failure. Radiation damage is fundamentally due to freely migrating point defects produced in collision cascades. A reduction in the freely migrating point defect population should, then, reduce radiation damage and increase component lifetime. The addition of oversized solute atoms such as Zr or Hf to 316SS, a common structural material in reactors, is expected to reduce point defect population through a trapping mechanism that enhances recombination. The mechanism, however, requires a strong binding energy between the oversized solute atom and vacancies in order for the mechanism to significantly reduce the defect population. Experimental measurements of this binding energy are unavailable, but can be determined with atomistic calculations. Ab initio methods are used here to determine binding energies and atomic volumes of either Hf or Zr oversized solutes with vacancies in a face-centered cubic Fe matrix. The binding energies are then used to parameterize a kinetic rate-theory model, which is used here to calculate radiation-induced segregation (RIS). The calculated values of RIS are then compared to experimental measurements to benchmark the calculations and offer insight into the proposed point defect trapping mechanism.
12:45 PM - JJ2.7/GG7.7
Behaviour of Irradiated hcp Zirconium Coupling Molecular Dynamics and Monte Carlo Simulations.
Cristina Arevalo 1 , Maria Caturla 1 , José Perlado 2
1 Instituto de Fusión Nuclear, UPM, madrid Spain, 2 Física Aplicada, Universidad de Alicante, Alicante Spain
Show AbstractRadiation damage in hexagonal-close-packed (hcp) metals is different from face-centred cubic (fcc) or body-centred (bcc) metals. The experimental study of point defect clustering in hcp metals is dominated by a consideration of the geometry of the hcp lattice and lattice parameters ratio (c/a). Because of this crystallographic anisotropy, defect anisotropic diffusion is expected, that is jump distances and jump rates depend on jump directions. We have studied irradiation of hcp α-Zirconium under different conditions with a kinetic Monte Carlo model. The initial cascade damage from Molecular Dynamics simulations produced by recoils from 10keV to 25 keV energies at 600K which is the operation temperature of the reactor have been followed for times of hours. The evolution of the microstructure under irradiation conditions of dose rate of 10-6 dpa/s and 600K has been studied until a final dose of 0.5 dpa. Using these calculations as the starting point we have compared them studying the influence of several parameters as dose rate, simulation box, grain size, bias and mobility of interstitials in the accumulation results.
JJ3
Session Chairs
Alfredo Caro
Francois Willaime
Tuesday PM, November 28, 2006
Berkeley (Sheraton)
2:30 PM - JJ3.1
DDD Simulations of Interactions Between Y2O3 Oxide Particles and an Edge Dislocation in ODS Materials.
Botond Bako 1 , Daniel Weygand 2 , Maria Samaras 1
1 , Paul Scherrer Institute, Villigen PSI Switzerland, 2 , IZBS Universitaet Karlsruhe, Karlsruhe Germany
Show Abstract2:45 PM - JJ3.2
A Dynamical Model for Clear Channel Formation in Irradiated Materials.
Silvester Noronha 1 , G. Ananthakrishna 2 , Nasr Ghoniem 1
1 Mechanical & Aerospace Engineering, UCLA, Los Angeles, California, United States, 2 Materials Research Centre, Indian Institute of Science, Bangalore India
Show Abstract3:00 PM - **JJ3.3
Dislocation Dynamics in fcc Metals.
Enrique Martinez 1 , Jaime Marian 1 , Athanasios Arsenlis 1 , Maximo Victoria 2 1 , J. Manuel Perlado 2
1 Chemistry and Materials Science Directorate, Lawrence Livermore National Laboratory, Livermore, California, United States, 2 Instituto de Fusion Nuclear, Universidad Politecnica de Madrid, Madrid, Madrid, Spain
Show Abstract3:30 PM - **JJ3.4
Fracture Properties a Ferritic Alloy and of a Tempered Martensitic Steel in the Transition.
Philippe Spatig 1 , Raul Bonade 1 , Pablo Mueller 1
1 CRPP, EPFL, Villigen Switzerland
Show Abstract4:45 PM - JJ3.6
Relationship Between the Magnetic and the Structural Properties of FeCr alloys: XAFS and PEEM Investigations.
Annick Froideval 1 , Maria Samaras 1 , Max Victoria 1 2 3 , Wolfgang Hoffelner 1
1 Nuclear Energy and Safety department, Paul Scherrer Institute, Villigen Switzerland, 2 , Lawrence Livermore National Laboratory, Livermore, California, United States, 3 Instituto de Fusion Nuclear, Polytechnic University of Madrid, Madrid Spain
Show Abstract5:00 PM - **JJ3.7
High Burnup Fuel Cladding Materials R&D for Water-cooling Nuclear Power Plants -- Nano-sized Oxide Dispersion Strengthening Steels--
Akihiko Kimura 1 , Hang-Sik Cho 2 , Naoki Toda 2 , Ryuta Kasada 1 , Hirotatsu Kishimoto 1 , Noriyuki Iwata 1 , Shigeharu Ukai 3 , Masayuki Fujiwara 4
1 Institute of Advanced Energy, Kyoto University, Kyoto Japan, 2 Graduate School of Energy Science, Kyoto University, Kyoto Japan, 3 Oarai Engineering Center, Japan Nuclear Cycle Development Institute, Ibaraki Japan, 4 , Kobelco Research Institute, Kobe Japan
Show Abstract5:30 PM - JJ3.8
Neutron Irradiation Resistance of RAFM Steels.
Ermile Gaganidze 1 , Bernhard Dafferner 1 , Jarir Aktaa 1
1 Institute for Materials Research II, Forschungszentrum Karlsruhe GmbH, Eggenstein-Leopoldshafen Germany
Show AbstractNeutron irradiation-induced embrittlement and hardening of the reduced-activation ferritic/martensitic (RAFM) steel EUROFER97 was studied under different heat treatment conditions. Irradiation was performed in the Petten High Flux Reactor within the HFR Phase-IIb (SPICE) irradiation project up to 16.3 dpa (in steel) and at different irradiation temperatures (250, 300, 350, 400, and 450°C). The embrittlement behaviour and hardening are investigated by instrumented Charpy-V tests with subsize specimens (KLST-type).The embrittlement behaviour of EUROFER 97 is compared with the results on international reference steels (F82H-MOD, OPTIFER-Ia, GA3X and MANET-I) included in the SPICE project. Analysis of embrittlement of RAFM steels within a proper model in terms of the parameter C=ΔDBTT/Δσ indicates hardening-dominated embrittlement at irradiation temperatures below 350°C with 0.17 ≤ C ≤ 0.53°C/MPa. Scattering of C at irradiation temperatures above 400°C indicates non hardening embrittlement. A role of He in a process of non hardening embrittlement is investigated in EUROFER 97 based steels, that are doped with different contents of natural boron and the separated 10B-isotope (0.008-0.112 wt.%).In view of qualification of the 2nd batch of EUROFER97-2 as well as novel tungsten and ODS alloys, a need of performance of a new irradiation programme will be discussed. For this purpose extension of the irradiation temperature spectrum towards higher temperatures, i.e. at 550 and 650 °C is proposed. Novel testing methods for determination of quasi-static fracture toughness as well as novel specimen geometries will be also presented.
5:45 PM - JJ3.9
Strength Analysis of Reduced Activation Ferritic/Martensitic Steel by Indentation Test.
Motoki Nakajima 1 , Shin-ichi Komazaki 1 , Mikio Fujiwara 1 , Yutaka Kohno 1 , Kiyoyuki Shiba 2 , Akira Kohyama 3
1 Department of Materials Science and Engineering, Muroran Institute of Technology, Muroran Japan, 2 , Japan Atomic Energy Agency, Tokai Japan, 3 Institute of Advanced Energy, Kyoto University, Kyoto Japan
Show Abstract
Symposium Organizers
Jarir Aktaa Forschungszentrum Karlsruhe GmbH
Maria Samaras Paul Scherrer Institute
Magdalena Serrano de Caro Lawrence Livermore National Laboratory
Maximo Victoria Polytechnic University of Madrid
Brian Wirth University of California-Berkeley
JJ4
Session Chairs
Mikhail Sokolov
Naoki Soneda
Wednesday AM, November 29, 2006
Berkeley (Sheraton)
9:30 AM - **JJ4.1
In Situ Transmission Electron Microscopy Study of Irradiation Effects in Refractory Ceramics for Application in Advanced Nuclear Systems.
Lumin Wang 1
1 Nuclear Engineering & Radiological Sciences and Materials Science & Engineering, University of Michigan, Ann Arbor, Michigan, United States
Show AbstractRefractory ceramics, such as silicon carbide, spinel and yttria stabilized zirconia (YSZ) are important candidates for components including inert fuel matrix in advanced nuclear reactors that may suffer high radiation dose in service. Extensive study of irradiation effects in these materials has been carried out with in situ Transmission Electron Microscopy (TEM) using the HVEM- or IVEM-Tandem Facility at Argonne National Laboratory that links the electron microscope with an ion beam accelerator. The irradiations were conducted in a wide range of temperatures using a wide variety of ion sources, including Ne, Ar, Ke, Xe, Cs, Sr and I in the energy range of 200 to 1500 keV to simulate the situation of high displacement damage and fission product incorporation in these materials. Cross-sectional TEM study have also been conducted in irradiated bulk samples to make sure that the results observed from pre-thinned TEM specimens are valid. The effects observed including dislocation loop and cavity formation in YSZ and spinel, cation disordering (in spinel) and amorphization (in SiC) at relatively low temperatures. At high temperatures, Sr precipitate out from YSZ in a secondary strontium zirconate solid phase while the volatile fission products, such as Kr, Xe, I and Cs tend to precipitate out as large gas bubbles. The gas bubbles accumulate at the phase boundaries in the YSZ/spinel composite. This talk will review the results of previous experimental results and discuss the implications of these results for the potential application in advanced nuclear systems.
10:00 AM - JJ4.2
Valence-dependent Analytic Bond-order Potential for Transition Metals.
Ralf Drautz 1 , David Pettifor 1
1 Department of Materials, University of Oxford, Oxford United Kingdom
Show Abstract10:15 AM - **JJ4.3
Modeling Radiation Damage in FeCr Alloys.
Alfredo Caro 1 , E. Lopasso 2 , M. Caro 1 , B. Sadigh 1 , D. Crowson 3 , S. Srivilliputhur 4
1 Materials Science, Lawrence Livermore National Laboratory, Livermore, California, United States, 2 , Centro Atomico Bariloche, Bariloche Argentina, 3 , Virginia Polytechnic Institute and State University, Blacksburg, Virginia, United States, 4 , Los Alamos National Laboratory, Los Alamos, New Mexico, United States
Show Abstract10:45 AM - JJ4.4
Thermo-mechanical Modelling of Pebble Beds in Fusion Blankets and its Implementation with Return-Mapping Algorithm.
Yixiang Gan 1 , Marc Kamlah 1
1 IMF II, Forschungszentrum Karlsruhe, Karlsruhe Germany
Show AbstractIn this investigation, a thermo-mechanical model of pebble beds is adopted (Hofer and Kamlah, 2005) and developed based on experiments at Forschungszentrum Karlsruhe (FZK, Reimann et al. 2006). The framework of the present material model is composed of a non-linear elastic law, the Drucker-Prager-Cap theory, a modified creep law. Furthermore, the volumetric inelastic strain dependent thermal conductivity of beryllium pebble beds is taken into account and full thermo-mechanical coupling is considered. By analyzing the deformation mechanism of the oedometric experiments, the identification method is developed to determine the set of material parameters, including the temperature dependent hardening law.The research shows that the Drucker-Prager-Cap model implemented in ABAQUS can not fulfill the requirements of both the prediction of large creep strains and the hardening behaviours caused by creep, which are important during the application of pebble beds in fusion blankets, so UMAT and UMATHT routines are used to re-implement the present thermo-mechanical model in ABAQUS. The elastic predictor radial return mapping algorithm is used to solve the non-associated plasticity iteratively, and a proper tangent stiffness matrix is obtained for the cost-efficiency in calculation. The explicit creep mechanism is adopted for the prediction of time-dependent behaviours. Furthermore, the thermo-mechanical interactions are implemented in UMATHT routine for the coupling analysis.The oedometric compression tests and creep tests of pebble beds under different temperatures are simulated by the present UMAT and UMATHT routines, and the comparisons between the simulation and the experiments are made. The applications on the large scale experiments are applied to show the feasibility of the present material model.
11:30 AM - JJ4.5
Studying Role of Magnetism in Defect Mobilities of Fe using MD Simulations.
Maria Samaras 1 , Max Victoria 2 3
1 , Paul Scherrer Institute, Villigen Switzerland, 2 Nuclear Fusion, Universidad Politecnica de Madrid, Madrid Spain, 3 , Lawrence Livermore National Laboratory, Livermore, California, United States
Show Abstract11:45 AM - JJ4.6
Including the Effects of Electronic Excitations and Electron-Phonon Coupling in Cascade Simulations.
Dorothy Duffy 1 2 , Alexis Rutherford 1
1 Physics and Astronomy, UCL , London United Kingdom, 2 Culham Science Centre, EURATOM/UKAEA Fusion Association, Oxfordshire United Kingdom
Show Abstract12:00 PM - JJ4.7
Stable one-dimensional Migration of Radiation Defects in Tungsten.
Sergei Dudarev 1 3 , Peter Derlet 2 , Duc Nguyen-Manh 1
1 Theory and Modelling, EURATOM/UKAEA Fusion Association, Oxfordshire United Kingdom, 3 Physics, Imperial College, London United Kingdom, 2 ASQ/NUM, Materials Science & Simulation, Paul Scherrer Institute, Villigen PSI Switzerland
Show Abstract12:15 PM - JJ4.8
Modelling of Magnetic Defect Structures in Ferro-magnetic bcc Fe.
Peter Derlet 1 , Peter Van Zwol 1 , Helena Van Swygenhoven 1 , Sergei Dudarev 2
1 ASQ/NUM - Materials Science & Simulation, Paul Scherrer Institute, PSI-Villigen Switzerland, 2 EURATOM/UKAEA Fusion Association, Culham Science Centre, Oxfordshire United Kingdom
Show Abstract12:30 PM - **JJ4.9
Irradiation Behavior of Metallic Nuclear Reactor Fuel.
Gerard Hofman 1 , Yeon Soo Kim 1
1 Nuclear Engineering, Argonne National Laboratory, Argonne, Illinois, United States
Show AbstractWednesday PM, November 29, 2006
Berkeley (Sheraton)
2:30 PM - JJ5.1
Radiation Damage Evolution due to Helium Implantationin Nb-Cu Nanolaminate Composites.
Amit Misra 1 , Michael Demkowicz 2 , Y. Wang 2 , Richard Hoagland 2
1 MPA-CINT, Los Alamos National Laboratory, Los Alamos, New Mexico, United States, 2 MST-SPR, Los Alamos National Lab, Los Alamos, New Mexico, United States
Show Abstract2:45 PM - JJ5.2
Thermal Helium Desorption of Helium-Implanted Iron.
Donghua Xu 1 , Brian Wirth 1
1 Nuclear Engineering, University of California, Berkeley, Berkeley, California, United States
Show AbstractFerritic martensitic steels will experience severe irradiation induced degradation of many important performance sustaining mechanical properties in fusion environments, driven by simultaneous production of displacement defects and high concentrations of helium. A key issue is coupled transport and fate of all defect, gas and solute species. In this work, we focus on determining the mechanisms of helium interaction, trapping and migration mechanisms in Fe and ferritic alloys. Thermal helium desorption spectroscopy (THDS) measurements have been performed on nominally pure iron specimens, implanted with helium under different conditions. Helium-implantations were performed at energies from 5 to 100 keV and at doses from 1E11 /cm2 to 1E15 /cm2 on iron specimens as a function of grain size and dislocation density. The experimental results yield the desorption temperature, the activation enthalpy for desorption, the attempt frequency for desorption, and an indication of the types of defects from which helium is desorbing. The experimental results are compared with recent molecular dynamics and kinetic Monte Carlo simulations on the energetics and migration mechanisms of helium, and its interactions with point defect clusters and extended defects in iron.
3:00 PM - JJ5.3
The Effects of Helium and Hydrogen in Irradiated Body-Centered Cubic Iron.
Maria Okuniewski 1 , Chaitanya Deo 2 , Marc Weber 3 , Farida Selim 3 , Kelvin Lynn 3 , Srinivasan Srivilliputhur 2 , Stuart Maloy 2 , Michael Baskes 2 , Michael James 4 , James Stubbins 1
1 Nuclear, Plasma, and Radiological Engineering, University of Illinois at Urbana-Champaign, Urbana, Illinois, United States, 2 MST-8, Los Alamos National Laboratory, Los Alamos, New Mexico, United States, 3 Center for Materials Research, Washington State University, Pullman, Washington, United States, 4 D-5, Los Alamos National Laboratory, Los Alamos, New Mexico, United States
Show AbstractAccelerator driven systems, generation IV reactors, and fusion systems are considering the use of ferritic and ferritic-martensitic steels for structural materials. These steels were selected because of their resistance to void swelling, irradiation creep, and helium (He) and hydrogen (H) embrittlement at higher temperatures. During the operation of the reactor, these materials will be subjected to both irradiation displacement damage, as well as the generation of H and He through (n,p) and (n,α) transmutation reactions, respectively. These transmutation gases and irradiation damage have a significant impact on the resultant microstructure and material properties. This topic has been the focus many experimental and modeling studies over the past twenty years. The current work is aimed at quantifying these effects in ways that were not possible in earlier studies. This is accomplished through a systematic group of coordinated computational modeling and experiments. The modeling approach employs molecular dynamics (MD) and kinetic Monte Carlo (kMC) simulations to study the dynamic evolution of He and defect clusters in body-centered cubic (BCC) iron (Fe). The defect configuration information from MD is then used in the kMC simulations to study point defect diffusion and clustering. The kMC model follows the transport and evolution of major defect entities in the material including, interstitial and substitutional He, Fe self-interstitial atoms, vacancies, vacancy-clusters, and sinks for the trapping of point defects (dislocations and grain boundaries). The time evolution is tracked for defect clustering and bubble formation as a function of irradiation conditions, times and temperatures. The experimental studies utilized ion implantation in single crystal BCC Fe to simulate the radiation damage processes over a range of He/dpa values and dose levels. The resulting microstructures are characterized using positron annihilation spectroscopy and TEM. The experimental results are compared to the computational findings.
3:15 PM - JJ5.4
Dissolution of Helium in Iron and other bcc Metals from First Principles.
Francois Willaime 1 , Chu Chun Fu 1
1 SRMP, CEA/Saclay, Gif-sur-Yvette France
Show Abstract3:30 PM - **JJ5.5
Structural Materials for Fusion Reactors Radiation Effects and Major Issues.
Jean-Louis Boutard 1
1 , EFDA-CSU Garching, Garching bei München Germany
Show Abstract4:30 PM - **JJ5.6
Ductility of Tungsten Alloys: A Critical Parameter for Fusion Application.
Mario Faleschini 1 , Herbert Kreuzer 1 , Reinhard Pippan 1
1 Erich Schmid Institute of Materials Science, Austrian Academy of Sciences, Leoben Austria
Show Abstract5:00 PM - JJ5.7
Finite Element Analysis of Functionally Graded Tungsten/Steel Joints.
Mohamed-Karim Hajji 1 , Jarir Aktaa 1
1 , Forschungszentrun Karlsruhe, Eggenstein-Leopoldshafen Germany
Show Abstract5:15 PM - JJ5.8
Electron-beam Induced Recrystallization in Amorphous Apatite.
In-Tae Bae 1 , Yanwen Zhang 1 , William Weber 1
1 , Pacific Northwest National Laboratory, Richland, Washington, United States
Show Abstract5:30 PM - **JJ5.9
Irradiation Assisted Grain Boundary Segregation in Steels.
Roy Faulkner 1
1 IPTME, Loughborough University, Loughborough, Leics, United Kingdom
Show AbstractJJ6: Poster Session
Session Chairs
Thursday AM, November 30, 2006
Exhibition Hall D (Hynes)
9:00 PM - JJ6.1
Structural Study of SiC Nanoparticles Grown by Inductively Coupled Plasma and Laser Pyrolysis for Nanostructured Ceramics Elaboration.
Yann Leconte 1 2 , Marc Leparoux 2 , Nathalie Herlin-Boime 1 , Stephan Siegman 2 , Lukas Rohr 2 , Cécile Reynaud 1
1 dsm/drecam/spam, CEA, Gif sur Yvette France, 2 Laboratory for Materials Technology, EMPA, Thun Switzerland
Show Abstract9:00 PM - JJ6.2
Molecular Dynamics Simulations of Defect Structures in UO2.
Antonino Romano 1 , Maria Samaras 2 , Martin Zimmerman 1
1 Laboratory for Reactor Physics and Systems Behaviour, Paul Scherrer Institut, Villigen-PSI, Aargau, Switzerland, 2 Laboratory for Materials Behaviour, Paul Scherrer Institut, Villigen-PSI, Aargau, Switzerland
Show Abstract9:00 PM - JJ6.3
Fracture Toughness and Tensile Properties of Two Mo-Re Alloys.
Mikhail Sokolov 1
1 , ORNL, Oak Ridge, Tennessee, United States
Show Abstract9:00 PM - JJ6.4
Fabrication and Characterization of Thin Sputtered Commixed Al/Gd Films.
Javier Jaquez 1 , Hongwei Xu 1 , Abbas Nikroo 1
1 Inertial Confinement Fusion, General Atomics, San Diego, California, United States
Show Abstract9:00 PM - JJ6.6
The All Boron Carbide Diode Neutron Detector: Experiment and Theory.
Ildar Sabirianov 1 2 , Robert Fairchild 3 , Jennifer Brand 1 2
1 College of Engineering, UNL, Lincoln, Nebraska, United States, 2 Nebraska Center for Materials and Nanoscience, UNL, Lincoln, Nebraska, United States, 3 Physics and Astrnomy, Nebraska Wesleyan University, Lincoln, Nebraska, United States
Show Abstract9:00 PM - JJ6.7
The Role of Grain Boundary/Interstitial Interactions in Thin Film Stress Evolution: Insights from Atomic Simulations.
Stephen Foiles 1 , Chun-Wei Pao 2 , Edmund Webb 1 , David Srolovitz 3 , Jerrold Floro 1
1 , Sandia National Laboratories, Albuquerque, New Mexico, United States, 2 Mechanical and Aerospace Engineering, Princeton University, Princeton, New Jersey, United States, 3 , Yeshiva University, New York, New York, United States
Show Abstract
Symposium Organizers
Jarir Aktaa Forschungszentrum Karlsruhe GmbH
Maria Samaras Paul Scherrer Institute
Magdalena Serrano de Caro Lawrence Livermore National Laboratory
Maximo Victoria Polytechnic University of Madrid
Brian Wirth University of California-Berkeley
Thursday AM, November 30, 2006
Berkeley (Sheraton)
9:30 AM - JJ7.1
Microstructure and High Temperature Properties of Ni-W and Ni-W-Cr Alloys.
Rafael Cury 1 , Thierry Auger 1 , Jean-Pierre Chevalier 2 1
1 CECM, CNRS, Vitry France, 2 Chaire of industrial materials, CNAM, Paris France
Show Abstract9:45 AM - JJ7.2
A Mechanical Study of T91 and 316L Embrittlement by Liquid Lead-bismuth Eutectic.
Zehoua Hamouche 1 , Thierry Auger 1 , Ivan Guillot 1
1 , CECM-CNRS, Vitry-sur-Seine France
Show Abstract10:00 AM - **JJ7.3
Modeling of Neutron Irradiation Embrittlement of Reactor Pressure Vessel Steels and Effect of Dose Rate.
Naoki Soneda 1 , K. Dohi 1 , A. Nomoto 1 , K. Nishida 1 , S. Ishino 1
1 Materials Science Research Lab, Central Research Institute of Electrical Power Industry, Tokyo Japan
Show Abstract10:30 AM - JJ7.4
Bubble Density Dependent Functionals to Describe Deformation and Stress Equilibrium Evolution for In-Reactor Nuclear Fuel Materials.
Ray Stout 1
1 , RhoBetaSigma Affaires, Livermore, California, United States
Show AbstractThe evolution of dense sets of discrete bubble species(order 10E+12 to 10E+18 bubbles per cc) in nuclear materials is described in terms of a generic Boltzmann transport equation[1]. A bubble species is identified by the physical attributes of radius, gas content, and their rates with respect to time. A bubble density function is defined as the number of bubbles per unit bubble species per unit spatial volume at an arbitrary time. The in-reactor nuclear fuel material responses for deformation and stress equilibrium are bubble density dependent, as the radius and gas content of a bubble species will evolve in time and space at rates compatible with the local fission rate, temperature, and stress state. The use of a generic Boltzmann equation for bubble density evolution is a direct analog to that used in gas kinetics; for which the evolution of dense sets of discrete gas atoms/molecules(order 10E+21 to 10E+23 per cc) is described in terms of the Boltzmann transport equation. In gas kinetics, Boltzmann’s equation is a conservation expression for each gas species, and in gas kinetics a gas species is identified by the physical attribute of velocity vector. For nuclear fuel materials, a spatial volume will contain material spatial subsets that enclose dense sets of discrete bubble species. Thus, the spatial volume of material is a non-contiguous material; and is not a continuum for applied mathematical operations of derivatives and integrations[2, 3, 4]. By using vector path integral concepts for arbitrary spatial paths between any two arbitrary and finitely separated spatial points, two stochastic vectors are derived that decompose the spatial volume into material and bubble spatial domains at any arbitrary time. This vector decomposition is used to derive a relative deformation functional that describes relative deformations, material and bubble density dependent, during the evolution of the bubble density function. From the deformation functional, an Eulerian strain measure is derived and is a bubble density dependent functional. For small strain approximations, the Eulerian strain tensor is separable into a material strain tensor and a bubble density strain tensor; in general this Eulerian strain tensor is not separable. The stress equilibrium derivation includes terms for material stress and for gas pressure stresses of discrete bubble species; it is also a functional that depends on bubble density evolution. References1. Boltzmann, Ludwig[1964]: Lectures on Gas Theory, translated by S. Brush, University of CalPress, CA.2. Shilov, G.E., and B.L. Gurevich[1977]: Integal, Measure,& Derivative: A Unified Approach, Dover Publications, Inc.(NY).3. Riesz, F. and B. Sz.-Nagy[1990]: Functional Analysis, Dover Publications edition of Ungar-Publication Co.(NY) 1955 edition.4. Stout, R.B.[2006]: Stochastic Deformations and Bubble Density Evolution in Nuclear Materials, RBSAffaires Rpt 00016(Jun06), Livermore, CA.
10:45 AM - JJ7.5
Atomistic Simulations of Displacement Cascades in Fused Silica: It is Compared with Different Concentration of H in the Bulk.
Fernando Mota 1 , Maria José Caturla 3 , J.Manuel Perlado 1 , Joaquín Molla 2 , Angel Ibarra 2
1 Instituto de Fusión Nuclear, E.T.S.I.I Universidad Politenica de Madrid, Madrid , Madrid, Spain, 3 Fisica Aplicada, Universidad de Alicante, Alicante, Alicante, Spain, 2 Materiales para la fusion, CIEMAT, Madrid, Madrid, Spain
Show Abstract11:30 AM - JJ7.6
Microstructure and Mechanical Properties of n-irradiated Fe-Cr Model Alloys.
Milena Matijasevic 1 , Abderrahim Almazouzi 1
1 Reactor materials research, SCK-CEN, Mol Belgium
Show Abstract11:45 AM - JJ7.7
Use of RHEPP-1 for Long-term Exposure of Reactor First-Wall Materials to Energetic Pulsed Ions.
Timothy Renk 1 , Paula Provencio 1 , Tina Tanaka 1 , Craig Olson 1
1 , Sandia National Laboratories, Albuquerque, New Mexico, United States
Show AbstractIn future Laser Inertial Fusion Energy (IFE) power plants, MeV-ions will impinge on the chamber wall at high fluence (up to 20 J/cm2) and at rates up to 10 Hz. The effects on proposed chamber materials subjected to long-term pulsed ion beam exposure (up to 2000 pulses) are being investigated on the 800 kV RHEPP-1 facility at Sandia National Laboratories. For these studies, beams of helium and nitrogen ions (200-400 ns pulsewidth on target, at fluences up to and exceeding ablation threshold) are directed onto samples of candidate wall armor materials such as tungsten, tungsten alloy and graphite. Samples are exposed at room temperature, or up to 600° C, and examined with SEM and XTEM as well as with compositional diagnostics such as XPS and EDS. Exposed samples are also measured for weight loss. It is important to determine the threshold for surface morphology changes in these materials. Any developing surface morphology may be mechanically unstable, i.e. may lead to loss of material by exfoliation or similar mechanisms. In addition, surface microcracks and roughness may act as initiation sites for fatigue cracks which may then propagate below the surface. In the case of tungsten, it appears that the various forms investigated (powder-metallurgy (PM), single-crystal and fine-grain) are unaffected by repeated exposures below about 1 J/cm2 (nitrogen beam). SEM imaging at 15,000 magnification shows no evident micro-cracking. This corresponds to a peak surface temperature of 2,000 – 2,500K, depending upon if the sample is heated or not. Above this level, however, PM tungsten can evolve surface relief rapidly with pulse number. Relief is much less evident for the other forms of tungsten, and for tungsten deformed with grains perpendicular to the surface. This suggests that the PM tungsten (polycrystalline) behavior is dominated by the mechanical instability of the (large) grains exposed to ion bombardment. All of the exposed graphite showed changes in surface appearance and/or morphology, even at relatively low fluence. These results have implications for plasma-facing components in Magnetic Fusion Energy (MFE) devices as well. *Sandia is a multiprogram laboratory operated by Sandia Corporation, a Lockheed Martin Co., under US DOE Contract DE-AC04-94AL85000.
12:00 PM - JJ7.8
Brittle Fracture Behavior of Austenitic and Martensitic Steels Induced by Mercury at Room Temperature.
Aida Liliana Medina-Almazan 1 2 , Thierry Auger 1 , Philippe Bompard 2 , Dominique Gorse 1 , Colette Rey 2
1 CECM, CNRS, Vitry sur Seine France, 2 LMSSMat, CNRS-ECP, Chatenay-Malabry France
Show AbstractThis work is focused on the plastic deformation behavior of 316L (austenitic) and T91 (ferritic- martensitic) steels in contact with mercury at room temperature. Both steels are candidate structural materials for both nuclear systems dedicated to the transmutation of nuclear wastes, and especially spallation targets and also for generation IV reactor systems. Liquid Metal Embrittlement (LME) is the transition from a ductile to a brittle rupture when: 1) there is an intimate contact between the solid and the liquid metals (wetting) and 2) there is a plastic deformation that is required for liquid metal induced brittle crack initiation. It has been found that both, 316L and T91 steels, fail in a transgranular way, mostly by enhanced shear decohesion, in contact with mercury at room temperature. This embrittlement behavior is strain rate dependent. At strain rates where the embrittlement effect is present, it is observed that once the brittle crack is nucleated, there is surface strain localization. The evaluation of the small scale yield stress at the crack tip during its propagation, by a crack tip opening displacement (CTOD) analysis, shows that the work hardening seems limited by the action of mercury [1, 2]. The operating mechanism of this environmental fracture is not yet clarified.In order to improve the understanding of the mechanism of this modification in plastic deformation behavior, mechanical tests in air and in mercury using Compact Tension (CT) specimens have been coupled with Electron Backscatter Diffraction (EBSD) analysis. By using the quality and transgranular misorientation maps, EBSD permits the measurement of change in local deformation across the crack path. In that way, a comparison at the mesoscopic scale between the plastic deformation behavior in air and in mercury is done.This paper begins by a brief summary of the experimental results showing the embrittling effect of mercury on T91 and 316L steels. Preliminary results on CT specimens analyzed by EBSD will then be presented, with the perspective to propose a mesoscopic model of a LME transgranular fracture.[1]L. Medina-Almazán, D.Gorse, T. Auger, Brittle to ductile transition study of 316L steel in contact with Hg, to be submitted.[2]L. Medina-Almazán, T. Auger, D. Gorse, Wettability of iron base alloys by liquid metals and consequences on fracture processes, presented at Eurocorr 2005, Lisbon.
12:15 PM - JJ7.9
Effect of Minor Alloying Element on Dispersing Nano-particles in ODS Steel.
Somei Ohnuki 1 , Toshiyasu Nagai 1 , Yosuke Uchida 1 , Koichi Hamada 1 , Tamaki Shibayama 1 , Takanori Suda 1 , Naoaki Akasaka 2 , Shin-ichiro Yamashita 2 , Satoshi Ohtsuka 2 , Tsunemitsu Yoshitake 2
1 Grad. School of Engineering, Hokkaido University, Sapporo, Hokkaido, Japan, 2 Oarai-research center, JAEA, Oarai, Ibataki, Japan
Show AbstractFrom the irradiation resistance and high-temperature strength, oxide dispersion strengthened (ODS) ferritic steels are candidate materials for advanced and fusion reactors. For the development of advanced steels the key issue is to homogenize nano-particles into matrix. Recent studies have indicated that Ti addition can homogenize Y-Ti complex particles into ferrite matrix, but the reason of the effect of additional elements has not been clarified. In this model study, we focus on the effect of additional elements, such as IV and V families and other oxide formers, which can control potentially the distribution of the oxide particles. The materials used in this study were based on Fe-9Cr-Y2O3 alloys which were mechanical alloyed (MA) from the powder of Fe, Cr and Y2O3, which was added systematically with the element of Ti, Zr, Ta, V, Nb, Hf, Al, Si and others. Usually ODS fabrication process is required for hot extrusion, but we annealed up to 1150 C for simplify the microstructure. To evaluate the distribution of ODS particles; we used TEM equipped with EDS after electro-polishing or FIB techniques. (1) In the case of Si or Al addition, oxides were disappeared after MA process, which means Y2O3 and other elements should be in solution at non-equilibrium condition. Two types of oxides of Y2O3 and Al2O3 or SiO2 developed after the annealing at 850 C, but only complex oxides were developed after the annealing at 1150 C. This result suggests that the oxide formation is independent process for Y and Si or Al. (2) In the case of Ti addition, oxides also were disappeared after MA process, but developed after annealing at 1150 C. This means that Ti can stabilize complex oxides of Y and Ti, and enhance the fine distribution of the oxides comparing with simple Fe-9Cr-Y2O3 alloy.
12:30 PM - JJ7.10
Solid Micro-beads Hetero-structure Fuel for Ultra-high Temperature Applications.
Liviu Popa-Simil 1
1 , LAVM Inc., Los Alamos, New Mexico, United States
Show AbstractHigher conversion efficiencies require high operation temperatures that are difficult to obtain due to the actual thermo-physical properties of the nuclear fuels. The initial intrinsic thermal conductivity of the actual fuel pellets, mainly ceramics like structures (oxides, nitrides, carbides, MOX, beads) is low. The center of the pellet is near melting temperature while the cladding operation temperature has to be low. The fission products deposition and burnup effects are further dimming the thermal conductivity. More the cooling agent’s chemical reactivity increase with temperature is another main reasons of keeping the operation temperature low. The usage of a hetero-structure of solid fuel soaked into a drain fluid is increasing the thermal conductivity. Properly shaped beads structure drives to the possibility of preventing most of the fission products of being stored inside the fuel lattice deteriorating its properties, being drained outside the nuclear reactor. This changes inspired from the nature, makes the nuclear reactor resembling with a plant having self cleaning and curing properties while operating at higher temperatures. These higher temperatures are allowed by the increase of thermal conductivity by 3 to 20 times by immersing the fuel into liquid metal and by avoiding the its burnup dimming. Making the fission products trajectories to end in the drain liquid that is tolerant to the increased nuclear recoils and damage from the end of the range. Due to better thermal conductivity the temperature field differences inside the nuclear reactor becomes smaller, allowing the operating temperature to rise significantly without safety concerns. There is possible to continuously remove the fission products by smoothly circulating the drain liquid. The low flow is needed both to give time to fission products to end inside the reactor highly shielded volume the most of the short lives disintegration chains. There are several fissionable materials and drain liquids matching which assure high operation temperatures, and allows He cooling, and high temperatures gas turbines cycles. Depending on the chosen actinide-drain-liquid, cooling-liquid conversion efficiencies might be higher than 70%. The new concept on fission products continuous release and separation minimizes the waste and the total radioactivity stored inside the reactor to few weeks integrated operation amount remaining constant over the time. That makes the fuel’s remnant radioactivity lower by a factor >100 than the actual reactors level. The fuel reactivity might be controlled by poisoning and transmutation or by assuring specific reactive geometries which to allow a ultrahigh burnup without the need of over-criticality loads. The fuel’s deformability opens the way for interesting applications. The advantages of micro structured nuclear fuel are higher thermal conductivity, fission products removal, appropriate reactivity, higher efficiencies and longer fuel life.
12:45 PM - JJ7.11
Comparative Study of Structural Damage Under Irradiation in SiC Nanostructured and Conventional Ceramics.
Yann Leconte 1 , Isabelle Monnet 2 , Marc Levalois 4 , Magali Morales 4 , Xavier Portier 4 , Lionel Thomé 3 , Nathalie Herlin-Boime 1 , Cécile Reynaud 1
1 dsm/drecam/spam, CEA, Gif sur Yvette France, 2 dsm/drecam/ciril, CEA, Caen France, 4 SIFCOM, CNRS-Université de Caen, Caen France, 3 CSNSM, CNRS, Orsay France
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