Symposium : N
Nuclear materials IV
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| 09:00 | Welcome to Symposium N (C. Degueldre, Symposium Chair) | |
| 09:15 | Welcome to E-MRS (F. Priolo, E-MRS President) | |
| 09:30 | Introduction to Nuclear Materials Authors : Anton Moeslang1, Franck Carre2, Claude Degueldre3, Bill Lee4, Ronald Schramm5 1 KIT, Karlsruhe, D; 2 CEA, Saclay, F; 3 PSI, Villigen, CH; 4 Imperial College, London, UK; 5 NRG, Petten, NL Resume : Nuclear materials are studied for their specific utilisation under extreme temperature, pressure or irradiation environments. These materials act as barriers and their structural properties are investigated with emphasis on mechanical performances, durability, plasticity and stability. The symposium N includes sessions dealing with materials ranging from the fuel, structural components, fusion system components and waste form materials. Macro-properties such as thermodynamical, thermophysical and mechanical as well as microstructural analysis of these materials are discussed for example comparing properties prior and after irradiation. | 0 |
| Fusion materials : Moeslang (KIT) & Becquart (U Lille) | ||
| 10:00 | Effective mathematical models for fusion materials Authors : S L Dudarev, Culham Centre for Fusion Energy, United Kingdom Atomic Energy Authority, Abingdon, Oxfordshire OX14 3DB, UK Resume : The development of new materials for cost-effective fusion power generation has become one of the priority issues for the international magnetic fusion programme. While the criteria related to the stability of materials under 14 MeV fusion neutron irradiation, where candidate alloys and/or composite materials for the tritium breeding blanket are expected to retain mechanical strength, fracture toughness, and creep resistance over the period of 5 to 10 years, are broadly similar to those for the next generation fission reactor materials, there are significant differences, primarily associated with the different energy spectra of fusion and fission neutrons. Recent advances in mathematical modelling show that a synergetic approach, based on concepts drawn from theoretical condensed matter physics, novel computer simulation algorithms, and model validation experiments, offers a knowledge-based way forward in the development of fusion materials. This presentation focuses on a range of outstanding problems in the field of radiation stability of iron-based alloys and steels that have recently been resolved using new mathematical concepts and algorithms involving electronic, atomistic, mesoscopic (Langevin and Monte Carlo) and dislocation-based models. | I 1 |
| 10:30 | Coffee break | |
| 10:45 | Multiscale Study of Self-interstitial Behavior in Beryllium Authors : P.V. Vladimirov, Karlsruhe Institute of Technology, Karlsruhe, Germany; V.A. Borodin, RRC Kurchatov Institute, Moscow, Russia Resume : Beryllium is foreseen as neutron multiplier material for fusion blanket. Hence, radiation resistance and structural integrity of beryllium pebbles are of high importance. The mobility and interaction of radiation defects in Be are vital for the description of radiation induced microstructure evolution.
The work addresses the stability of self-interstitials and their clusters in hcp Be, as well as the clarification of interstitial diffusion pathways. The primary selection of non-trivial interstitial configurations and their diffusion paths was achieved using molecular dynamic with Be potential from [1]. The selected configurations and migration barriers were verified using first-principles code VASP.
We found that among the earlier predicted stable self-interstitial configurations [2] only two correspond to true energy minima (basal octahedral (BO) and crowdion (C)), while others are saddle points for diverse diffusion pathways. Crowdion is a shallow minimum that needs only 0.06 eV to decay into BO. We predict easy diffusion of self-interstitials both in basal planes (through basal split configuration with migration barrier of 0.12 eV) and between them (e.g. BO-C-BO with migration barrier of 0.27 eV).
Finally, we discuss the implications of the found SIA diffusion anisotropy for the radiation microstructure evolution.
[1] C. Björkas et al. J. Phys.: Condens. Matter 21 (2009) 445002.
[2] M.G. Ganchenkova, P.V. Vladimirov, and V.A. Borodin, J. Nucl. Mater., 386-388 (2009) 79. | I 2 |
| 11:00 | Pores and cracks in highly neutron irradiated beryllium Authors : V. Chakin(1), R. Rolli(2), A. Moeslang(1), P. Kurinskiy(1), W. Van Renterghem(3) (1) Karlsruhe Institute of Technology, FZK, IMF I, P.O. Box 3640, 76021 Karlsruhe, Germany (2) Karlsruhe Institute of Technology, FZK, IMF II, P.O. Box 3640, 76021 Karlsruhe, Germany (3) SCK CEN, Nuclear Materials Science Institute, Boeretang 200, B-2400 Mol, Belgium Resume : At present, beryllium is used as a reflector or moderator in research nuclear reactors. Prominent features of a reflector and moderator are a low irradiation temperature (50-70°С) and high limiting fast neutron fluence which corresponds to accumulation of helium up to 20000 аррm. In the future, beryllium will be used as a neutron multiplier in the blanket of DEMO where an irradiation temperature reaches 550-850°С, a damage dose of 80 dpa, helium accumulation of about 26000 аррm. Despite the difference in irradiation temperatures, high level of helium accumulation in a research nuclear reactor by the use of high-temperature annealing allows to estimate a degree of radiation damage of beryllium in the fusion reactor. In this study beryllium irradiated in the SM research nuclear reactor located in Dimitrovgrad, Russia, and in the BR2 nuclear reactor located in Mol, Belgium has been used. The irradiation was performed at a low temperature up to neutron fluence of 1.4х1023 cm-2 (En > 0.1 MeV) that corresponds to a damage dose of 65 dpa.
Beryllium for the use in a reactor is manufactured by powder metallurgy by hot pressing or hot extrusion methods. This always leads to formation of pores and beryllium oxide particles on grain boundaries already in initial state before irradiation. Under neutron irradiation, owing to anisotropic growth of grains in some crystallographic directions, evolution of grain boundary porosity occurs. In particular, the size of available pores increases and formation of new pores takes place in the same time. The assumption is that the grain boundary pore in irradiated beryllium can be as a crack germ. An increase of stretching strength from adjacent grains, which is produced owing to compression of the grains along the axis \\\"c\\\" of the beryllium HCP lattice, leads to the growth of a grain boundary crack. High-temperature annealing changes the process of crack formation in irradiated beryllium because the pores evaluate to gas bubbles owing to accumulation of radiogenic helium in high amount. The data on swelling, mechanical properties and evolution of microstructure examined by optical metallography and SEM in irradiated beryllium including high-temperature annealing is presented. | I 3 |
| 11:15 | Effect of Laser Shock Processing on Low Cycle Fatigue Behaviour of Reduced Activation Ferritic Martensitic Steel EUROFER 97 Authors : J.L. OCAÑA1, J.A. PORRO1, M. MORALES1, C. MOLPECERES1, L. RUIZ DE LARA1, P. FERNÁNDEZ2, R. VILA2, A. IBARRA2 1Centro Láser UPM. Universidad Politécnica de Madrid Campus Sur UPM. Edificio La Arboleda. Ctra. de Valencia, km. 7,300. 28031 Madrid. SPAIN. E-mail: jlocana@etsii.upm.es 2National Fusion Laboratory-CIEMAT. Fusion Technology Division. Avda. Complutense, 22. 28040 Madrid. SPAIN. Resume : Reduced activation ferritic martensitic (RAFM) steels are considered as the structural materials for the breeding blankets components of the future fusion reactors. They will be exposed to high temperatures and high levels of neutron irradiation, as well as to high mechanical and thermo-mechanical stresses under which the structural material integrity must be assured. The main advantages of RAFM steels are their low swelling, good surface capability and good compatibility with aqueous, gaseous and liquid metal coolants. However, nowadays there are some remaining issues related with the RAFM steels, especially their limited strength at high temperatures. RAFM steels show a drop in tensile strength at about 550ºC and a strong reduction in creep strength starting at 600ºC. In addition, softening occurs during cyclic loading, what may lead to maximum allowable loads much smaller than the limits predicted by the current design rules.
Because of this undesirable strength degradation, RAFM steels appear to be appropriate candidates for a significant enhancement of the strength stability (and hence improved life fatigue) through cold working provided by the Laser Shock Processing technique: The induction of a field of compressive residual stresses close to the facing surface after a cyclic cold plastic deformation process seems to be a promising issue regarding the possibility of life extension of critical components.
In the present paper, initial results on residual stresses fields induced on EUROFER 97 (the EU RAFM reference material) through the LSP technique are presented along with preliminary estimations on achieved low cycle fatigue life extension in typical specimens, both at ambient temperature and after several typical thermal cycles. | I 4 |
| 11:30 | Mechanical and Microstructural properties of EB Welded RAFM ODS-Eurofer Steel for Application in Helium Cooled Modular Divertor Concepts Authors : R. Lindau 1), U. Jäntsch 1), M. Klimenkov 1), and P. Norajitra 2) 1) Karlsruhe Institute of Technology, Institute for Materials Research I, 2) Karlsruhe Institute of Technology, Institute for Materials Research III P.O. Box 3640, 76021 Karlsruhe, Germany Resume : For specific blanket and divertor applications in future fusion power reactors a replacement of presently considered Reduced Activation Ferritic Martensitic (RAFM) steels as structural material by suitable oxide dispersion strengthened (ODS) ferritic martensitic or ferritic steels would allow a substantial increase of the operating temperature from ~550°C to about 650°C. In all cases appropriate joining technologies have to be developed. Diffusion welding techniques to perform similar and dissimilar joints have been studied successfully. Friction Stir Welding (FSW) has shown a good potential but application is limited due to geometrical restrictions and needs further development.
For the advanced helium-cooled modular divertor concept various joining techniques are required for joining the complex structural parts made of different materials. The electron beam welding process with its highly concentrated energy input has been investigated as a potential process to join divertor structures made of ODS Eurofer.
For this purpose, samples of ODS-Eurofer steel were welded using a PTR 150 kV/15kW EB welding facility. Two different post-weld heat treatments (PWHT) were applied to investigate their influence on the mechanical and microstructural properties of the welded joints. Miniaturised tensile specimens were used to determine the tensile behaviour in the temperature range between RT and 500°C. KLST specimens were used for Charpy impact tests.
The microstructure of the weld and heat affected zone as well as the fracture surface of the samples were examined using optical and scanning electron microscopy. The results of the transmission electron microscopical examinations are described more detailed in [1].
[1] M. Klimenkov et al.: Microstructural characterization of Electron Beam Welded RAFM ODS-Eurofer. (This conference) | I 5 |
| 11:45 | Microstructural characterization of Electron Beam Welded RAFM ODS-Eurofer Authors : M. Klimenkov, U. Jäntsch, R. Lindau and A. Möslang Karlsruhe Institute of Technology, Karlsruhe, Germany Resume : The use of Oxide Dispersion Strengthened (ODS) steels instead of presently considered conventional RAFM steels would enhance the efficiency of future fusion power plants by increasing the operating temperature by about 100 K to 650°C.
The advanced helium-cooled modular divertor concept requires various joining techniques to join the complex structural parts consisting of different structural materials. The electron beam welding process with its highly concentrated energy input has been investigated as a potential process to join divertor structures made of ODS-Eurofer. Strips of this ODS steel were EB welded under well defined conditions.
The microstructural characterization of the welding and heat-affected zone after different post-weld heat treatments was performed using scanning and transmission electron microscopy methods - Dualbeam SEM/FIB and (TEM). The SEM and low magnification TEM analysis show the changes of the microstructure in the welded area. In all specimens a grain coarsening has been observed. The TEM imaging and analytical TEM was applied to study the changes of the distribution of the carbide precipitates and ODS particles. A dissolution and agglomeration of the nano-sized ODS particles as well as the formation of complex structured yttrium containing larger particles has been observed. The obtained microstructural results can be correlated to the changes in the mechanical behaviour as reported in [1].
[1] R. Lindau et al.: Mechanical and Microstructural properties of EB Welded RAFM ODS-Eurofer Steel for Application in Helium Cooled Modular Divertor Concepts. (This conference) | I 6 |
| 12:00 | Fabrication and characterization of Spanish RAFM at a pilot plant scale Authors : D. Rodriguez1, A. Morán2, M. Serrano1, J. Belzunce2 1 Structural Materials Division. CIEMAT. Avda Complutense 22. 28040 Madrid 2 ITMA Parque Empresarial Principado de Asturias. C/Calafates Parcela L3.44. 33417 Aviles (Asturias) Resume : One of the main challenges for the realization of the future fusion reactor is the development and qualification of structural materials for first wall and breeding blanket. The fusion reactor application requires materials resistant to radiation damage, high temperature of operation, good corrosion properties and reduced activation potential.
Reduced Activaction Ferritic/Martenistic (RAFM) steels 9Cr are the main candidates for first wall and blanket of fusion reactors, due to their resistance to swelling and excellent thermal properties. These steels are based on the classical Cr-Mo steel but with a chemical composition modified in order to fulfil the low activation requirements, substituting the alloying elements with long decay times due to high activation by neutron irradiation. For this purpose the Mo, the Nb and Ni are avoiding or minimizing the radiological undesirable elements.
This paper shows the work carried out to develop at a pilot plant scale a RAFM steel with chemical composition and metallurgical properties very close to EUROFER steel. The steel was obtained in a Melting Pilot Plant. A High Vacuum Induction Melting Furnace (VIM) was used in order to control and avoid possible impurities and atmospheric pollution. Deformation and Quench Dilatometry has also been used to select the appropriate parameters for the thermomechanical and thermal treatments and the obtained results are also analyzed.
This project, corresponding to the T4 CONSOLIDER TECNO_FUS INGENIO 2010, is focussed basically on the evaluation of the microstructural and mechanical properties of a reduced activation ferritic/martensitic steel fabricated at a semi-industrial scale in Spain, which chemical composition fulfil or is very close to the compositional specifications and present similar metallurgical properties than the EUROFER steel.
Keywords: RAFM, Eurofer97, Fusion reactor. | I 7 |
| 12:15 | Lunch Break | |
| 13:15 | Helium clustering in tungsten Authors : P.E. Lhuillier 1,2*, T. Belhabib3, A.L. Thomann3, P. Desgardin1,2, T. Sauvage1,2, P. Brault3, Y. Tessier3, M.F. Barthe1,2 1 (CNRS, UPR3079 CEMHTI, 1D avenue de la Recherche Scientifique, 45071 Orléans cedex2, France) 2 (Université Orléans, Faculté des Sciences, Avenue du Parc Floral, BP 6749, 45067 Orléans cedex 2, France) 3 (GREMI Polytech’Orléans, 14 rue D’Issoudun, BP 6744, 45 067 Orléans Cedex 2, France) Contact : marie-france.barthe@cnrs-orleans.fr Phone : 33 2 38 25 54 29 Resume : Tungsten has been selected has the first wall material of the divertor of the ITER fusion reactor. Under operation in fusion reactor, the tungsten will be submitted to severe conditions, and in particular to high alpha particles bombardment flux. The consequence of helium implantation on the fate and lifetime of tungsten component is a major issue for qualifying the reliability of tungsten for fusion application. The aim of the study was to investigate the behavior of helium implanted in tungsten at low energy and low flux and its interaction with vacancy defects. Helium ions at 320 eV were introduced into tungsten by plasma immersion at the flux of 2.5×1018 ion/m-2/s-1 at different fluences. The helium retention properties of tungsten were investigated by Nuclear Reaction Analysis (NRA) and the effect of helium introduction on the tungsten lattice was examined by Positron Annihilation Spectroscopy. Helium implanted tungsten exhibits a very low retention which decreases with the implantation fluence. The desorption behavior of helium as a function of the temperature was measured by in-situ NRA. Desorption starts at very low temperature and suggests the shallow trapping of helium. SEM pictures of samples after post-implantation annealing shows sparse surface pores for annealing temperature higher than 973 K. The features are due to helium bubbles opening at the surface. The creation of helium-filled vacancy-type defects in the near surface layer (0.5 to ~20nm) has been detected by using Positron Annihilation Spectroscopy. To identify the routes of creation of such helium-filled defects, a rather low level of displacements was introduced in the samples by 12 MeV proton irradiation, prior to helium introduction. The results of Positron Annihilation Spectroscopy suggest that the early stage of creation of helium-filled vacancy clusters proceed from the ejection of self-interstitial atom from of growing helium interstitial cluster. | I 8 |
| 13:30 | Microstructural evolution of tungsten under irradiation : importance of the primary damage model Authors : C.S. Becquart1, A. De Backer1, U. Sarkar1, C. Domain1,2, C. Ortiz3, M. Hou4 1Unité Matériaux et Transformations, UMR 8207, Université de Lille 1, F-59655 Villeneuve d’Ascq Cedex, France 2EDF-R&D Département MMC, Les Renardières, F-77818 Moret sur Loing Cedex, France 3 Laboratorio de Fusión por Confinamiento Magnético, CIEMAT, E-28040 Madrid, Spain 4 Physique des Solides Irradiés et des Nanostructures CP234, Université Libre de Bruxelles, Bd du Triomphe, B-1050 Brussels, Belgium. Resume : We have used an ab initio based Object Kinetic Monte Carlo code to model the evolution of tungsten implanted with He atoms at different doses. The binding energies of He clusters, He-V complexes and self-interstitial-He complexes have been calculated within the density functional theory for small sizes, and extrapolated for larger sizes. The He atoms and the associated damage have been introduced either at random in the tungsten matrix or according to distributions provided by the Marlowe code, taking into account the crystalline nature of the matrix. The results indicate that in order to reproduce the experimental results, one has to model properly the implantation sequence. They furthermore underline the significant contribution of the crystalline nature of the matrix. | I 9 |
| 13:45 | Summary session I | |
| Gen IV materials : Carre (CEA) & Hernandez (CIEMAT) | ||
| 14:00 | European R&D activities on Structural Materials for Innovative Nuclear Systems Authors : C. Fazio1, A. Gessi2, D. Gomez-Briceno3, J. Henry4, L. Malerba5, M. Rieth1 1Karlsruhe Institute of Technology, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen, Germany 2Agenzia nazionale per le nuove tecnologie, l’energia e lo sviluppo economico sostenibile, CR Brasimone, 40032 Camugnano Bologna, Italy 3Centro de Investigaciones Energéticas, Medioambientales y Tecnológicas Avenida Complutense 22 28040 Madrid, Spain 4Commissariat à l’Énergie Atomique, Saclay, 91191 Gif sur Yvette cedex, France 5Studiecentrum voor Kernenergie - Centre D’Étude de L’Énergie Nucléaire, Boeretang 200, 2400 Mol Belgium Resume : In the European Frame the Sustainable Nuclear Energy Technology Platform has issued the Strategic Research Agenda, where the development and characterization of structural materials for innovative nuclear fission systems play a prominent role.
The European approach for the development of nuclear energy is based on three main pillars, i.e. improvement of LWR technology, the development of technologies to close the fuel cycle introducing Fast Neutron Reactors (FNR) and the technological development of high temperature reactors (V/HTR) for high temperature applications. These last two pillars foresee nuclear reactor designs with operating conditions as high temperatures, high thermal gradients, high burn-up and corrosive environments, which are not explored in the current nuclear industry. Accordingly a key issue becomes the selection of structural and clad materials, which should be able to perform in a safe manner under the envisaged operational and postulated transient conditions. The 9 Cr Ferritic / Martensitic (F/M) steels and Oxide Dispersion Strengthened (ODS) Fe-Cr alloys are indicated as reference materials for the core structures of the FNR and the 9 Cr F/M steel is also indicated as selected material for the vessel of the GFR and VHTR and as heat exchanger material for the SFR and LFR.
It has been considered that despite the industrial experience on 9Cr F/M steels, further data are needed to qualify their use in the in-service conditions planned for the nuclear systems. On the other hand the ODS alloys are a nearly unexplored domain, where knowledge-improvement is mandatory for their nuclear application.
Within the 7th European Framework Program a Cross Cutting Research Project has been launched to contribute to the development and qualification of these structural materials. The focus of this project, named GETMAT (GEneration IV and Transmutation MATerials), is on the procurement of reference materials (ODS and 9Cr steels), their welding/joining and their qualification, in terms of mechanical and corrosion resistance in appropriate conditions. Moreover, an extensive post irradiation examination (PIE) program, taking advantage of ongoing or completed irradiation experiments (see e.g. [1-2]) will contribute to the assessment of the two classes of materials, supported by multiscale modeling.
The aim of this work is to present the rationale, the scientific objectives and the structure of the GETMAT project.
References
[1] C. Fazio, et al. “The MEGAPIE-TEST Project: Supporting Research and Lessons Learned in First-of-a-Kind Spallation Target Technology”, Nucl Eng Des (2007).
[2]. C. Fazio, A Alamo, A. Almazouzi, D. Gomez-Briceno, F. Groeschel, F. Roelofs, P. Turroni and J. U. Knebel, „Assessment of Reference structural materials, heavy liquid metal technology and thermal-hydraulics for European waste transmutation ADS“,GLOBAL 2005 Proc., Tsukuba, Japan, Oct. 9-13, 2005. | II 1 |
| 14:30 | 3D atomistic modeling of microcrack propagation in iron Authors : V.A. Borodin(1) and P.V. Vladimirov(2) (1) RRC Kurchatov Institute, Kurchatov Sq., 1, 123182 Moscow, Russia (2) Karlsruhe Institute of Technology (KIT), 76131 Karlsruhe, Germany Resume : Ferritic-martensitic steels are broadly used in modern nuclear reactors and are among the primary candidates for future advanced fission (Generation IV) and fusion energy systems. Unfortunately, these steels have drawbacks, such as the low-temperature brittleness. Radiation-induced microstructure is known to shift the ductile-to-brittle transition temperature well above the room temperature, implying possible risks for the integrity of reactor components. Various kinds of steel manufacturing (e.g. oxide dispersion strengthening with the aim of improving high-temperature creep resistance) can also affect fracture resistance. A realistic estimate of possible risks for integrity failure requires clear understanding of the microscopic processes occurring during the crack initiation and development in the radiation environment.
We present the results of molecular dynamics (MD) modeling of microcrack propagation in iron. We demonstrate that in contrast to most of the common views, the brittle crack in iron propagates by the emission of slip bands from the crack tips. The 3D nature of MD model allows us to clarify the atomistic structure of slip bands and the detailed kinetics of slip band expansion and interaction with microstructural obstacles (such as vacancy clusters and microvoids). Based on the obtained results, we discuss the expected effects of radiation microstructure development in ferritic-martensitic steels on their fracture resistance. | II 2 |
| 14:45 | The effect of Ostwald ripening on the critical resolved shear stress Authors : Péter Dusán Ispánovity 1, Botond Bakó 2, Daniel Weygand 3, Maria Samaras 1 and Wolfgang Hoffelner 1 1 Paul Scherrer Institut, Villigen, Switzerland 2 Service de Recherches de Métallurgie Physique, CEA, Saclay, France 3 Karlsruhe Institute for Technology, izbs, Karlsruhe, Germany Resume : In Ni-base superalloys with low Al/Ti content (an example is the Inconel 617, which has been selected as a candidate material for GEN IV reactors) at medium temperatures the precipitation of γ\' particles is observed. These spherical precipitates have typically a radius of 20 nm and a volume fraction of about few percent. Since γ\' particles are strong obstacles to dislocation motion, this fine distribution leads to superior strength and creep resistance of these materials.
After the precipitation has been finished, if the material is still exposed to high temperature, the Ostwald ripening of the γ\' particles takes place. This leads to the degradation of mechanical properties. Here we study by simulation the coarsening phenomenon and determine the γ\' particle size distribution at different times. Then by discrete dislocation dynamics simulations we calculate the corresponding critical resolved shear stresses. This allows us to provide a more accurate life time assessment of such materials. | II 3 |
| 15:00 | Creep Properties of He-Irradiated and non-Irradiated TiAl at High Temperatures Authors : Per Magnusson, Jiachao Chen, Wolfgang Hoffelner Paul Scherrer Institute, Villigen, Switzerland Resume : Intermetallic titanium-aluminides (TiAl) are well accepted elevated temperature materials. In conventional applications their oxidation resistance limits the maximum operating temperature to about 800°C. Advanced Generation IV nuclear reactors operate in non-oxidizing environments. This could enlarge the applicability of these materials to temperatures above 1000°C. The evaluation of TiAl for applications in these reactors requires information on high temperature creep and knowledge of how irradiation affects creep properties.
Therefore, the creep properties of an intermetallic alloy Ti-46Al-2W-0.5Si (at%) including creep strain rate, time to 1% creep deformation and creep fracture were investigated using helium-irradiated samples and non-irradiated samples, at temperatures ranging from 700 °C to 950 °C and stresses ranging from 100 MPa to 400 MPa. The helium irradiation was performed using 24 MeV He-ions, homogeneously implanting the samples with up to 1000 appm (atomic parts per million) helium. Non-irradiated samples showed good creep resistance, irradiated samples showed a clear loss of ducitility and time to fracture. Crept and irradiated samples were studied with transmission electron microscopy, in order to determine the size, density and distribution of helium bubbles. | II 4 |
| 15:15 | Understanding the structure of Fe-Cr alloys using synchrotron based X-ray analysis Authors : A. Idhil, M. Samaras, C. Borca, M. Victoria, W. Hoffelner Paul Scherrer Institute, Switzerland Resume : Defects, clustering and segregation are issues which alter the structural and mechanical properties of a material. In ferritic steels, a candidate structural material for Gen IV and fusion reactors, these issues are of great importance in understanding the lifetime of the materials. High Cr ferritic steels naturally segregate. This means that depending on the Cr content, the material can be in a state of short-range order (below approx. 10%Cr) or contain clusters of Cr atoms above this concentration. In a nuclear environment defects can be produced as a result of irradiation damage. Electronic structure calculations have recently shown that these configurations are due to magnetism. To check such calculations which are at the fundamental building block of multiscale modeling schemes, model validation of these results are vital.
Several techniques have been used to characterize the microstructural and magnetic properties of several model FeCr alloys with up to 16% Cr content. In this contribution, X-ray diffraction and TEM techniques have been used to image and characterize the grain microstructure. Bulk magnetic properties at low temperature have been measured using a Physical Properties Measurement System. The local atomic environment and the presence of short range distribution around the Cr and Fe atoms have been measured using Extended X-ray Absorption Fine Structure (EXAFS). | II 5 |
| 15:30 | Coffee break | |
| 15:45 | The interaction of screw and edge dislocations with Cr precipitates in FeCr alloy: An atomistic study Authors : G. Bonny, D. Terentyev and L. Malerba SCK-CEN, Mol, Belgium Resume : Iron-chromium alloys are the base for ferritic and ferritic/martensitic steels, which have a wide range of applications as structural materials in aggressive high temperature environments, such as gas turbines in conventional power plants, or key components in future nuclear reactors. The appearance of Cr-rich precipitates (α\' prime) after thermal ageing or irradiation is a typical feature of high-Cr ferritic/martensitic steels. α\' particles, obstructing the motion of dislocations, are long known to be the cause of hardening and embrittlement, observed in steels and Fe-Cr binary alloys.
In this work, we consider the interaction of a screw dislocation and edge dislocation with Cr precipitates in a bcc Fe matrix using molecular dynamics techniques. We apply two different interatomic potentials: (i) one that has been widely used in the past five years in the literature to study primary damage and point defect properties and (ii) another one specifically accounting for the properties of the screw dislocation in the Fe-Cr system as obtained from the first principles calculations. The interaction mechanisms revealed using both potentials are described in detail and compared. The results obtained for the screw dislocation suggest that two principally different interaction mechanisms may operate depending on the interatomic model applied, while the results for the edge dislocation are very similar.
To conclude, the impact of the α-α\' separation on the change of the yield stress due to the pinning of dislocations by the Cr precipitates is discussed. | II 6 |
| 16:00 | CORROSION OF ADVANCED GAS REACTOR (AGR) FUEL CLADDING IN TRACE AQUEOUS ELECTROLYTE ENVIRONMENTS Authors : Chin H. Phuah (Imperial College London) Mary P. Ryan (Imperial College London) William E. Lee (Imperial College London) Resume : Grain boundary chromium depletion in steel is a radiation-induced effect which can impact on aqueous corrosion. This work is quantifying the corrosion potential of discrete grain boundary chromium depletion, produced by thermal sensitisation as a function of heating temperature and time, in 20Cr/25Ni/Nb austenitic stainless steel - the advanced gas reactor (AGR) fuel cladding material - in various electrolyte environments. To date annealed (1050°C for 2 hours, furnace cooled) and non-annealed AGR cladding have been sourced from Springfields Limited. Scanning electron microscope analysis indicates that the general microstructure of the annealed cladding comprises uniformly distributed 10μm grains along the x-y-z dimensions, and also occurrences of niobium-carbide precipitates. Transmission electron microscope analysis on Focus Ion Beam (FIB)-prepared specimens show consistent grain-grain boundary compositions: averaging 20.4±0.2%wt, 53.6±0.1%wt and 26.0±0.2%wt of chromium, iron and nickel, respectively. The non-annealed cladding, however, exhibits a 10.3±0.2% chromium depletion at the grain boundaries which is expected to decrease corrosion resistance. The corrosion potential of the non-annealed cladding is measured at +1.60V and +0.70V (less noble) in 0.001M and 0.01M (higher chloride concentration) of reagent grade sodium chloride solution, respectively, using the anodic polarization method in an Ag|AgCl reference electrode. | II 7 |
| 16:15 | Swelling of oxide dispersion strengthened (ODS) ferritic steel as a function of He irradiation temperature Authors : Pouchon, Manuel Alexandre / Nuclear Energy and Safety Dep., Paul Scherrer Institut, 5232 Villigen PSI, Switzerland Rebac, Tomislav / Nuclear Energy and Safety Dep., Paul Scherrer Institut, 5232 Villigen PSI, Switzerland Döbeli, Max / Institute of Particle Physics , Eidg. Techn. Hochschule, 8093 Zurich, Switzerland Degueldre, Claude / Nuclear Energy and Safety Dep., Paul Scherrer Institut, 5232 Villigen PSI, Switzerland Hoffelner, Wolfgang / , Nuclear Energy and Safety Dep., Paul Scherrer Institut, 5232 Villigen PSI, Switzerland Resume : Oxide dispersion strengthened (ODS) ferritic steels are promising candidates for high temperature applications in future nuclear reactor systems (GEN IV). The nano-sized oxides being dispersed in the matrix, represent important pinning points for dislocations, and therefore significantly enhance the temperature creep behavior. ODS steels are originally designed for conventional high temperature applications. The applicability in a nuclear reactor environment has therefore to be elaborated. Besides the classical exposures, as temperature, mechanical load and chemical environment, the irradiation can additionally modify the material. The irradiation hardening, irradiation creep and swelling are classical properties to be tested in this frame.
The paper concentrates on the latter property, the irradiation induced swelling. ODS steel samples (PM2000) are therefore exposed to a He ion beam with a varying energy (1-2 MeV) in order to generate an almost evenly damaged surface layer of 2.5 µm. Several of these irradiations are performed at increasing temperatures, in order to study the temperature dependency. Starting from room temperature with increasing temperature, first a reduction in the swelling is found, this can be allocated to the annealing of point defects. By further augmenting the temperature, an increase of the swelling is found; this can be associated with void formation. The paper will quantify these results and put them into relation with the theory. | II 8 |
| 16:30 | Influence of the surface finishing on the corrosion/oxidation behaviour of 316L and T91 in stagnant LBE Authors : F.J. Martín-Muñoz*, L. Soler-Crespo and D. Gómez-Briceño Structural Materials Division, CIEMAT, Building 30 Avda. Complutense 22, Madrid 28040, Spain Tel.: +34 913466730; Fax: +34 913466661 *e-mail: fco.javier@ciemat.es Resume : Lead bismuth eutectic (LBE) has been proposed as a candidate for coolant and/or as spallation target material of Accelerator Driven Systems (ADS) and as coolant for a Generation IV Reactor, the Lead-Cooled Fast Reactor (LFR).
However, a key problem for the use of LBE as coolant is the high solubility of the main elements of the structural steels in this heavy liquid metal at high temperatures. Corrosion by liquid metal can proceed via various processes: species dissolution, formation of the inter-metallic compounds at the steels/liquid metal interface, etc. The formation of an oxide scale on the steel surface can prevent the steel dissolution by liquid LBE. Surface finishing of the steels could play a significant role in the properties of the oxide layers formed on the steels, under the adequate conditions of oxygen content and temperature.
The objective of this investigation is to gain some insight into the influence of the surface finishing in the oxidation/corrosion behaviour of 316L and T91 steels. Specimens of both materials with different surface states were prepared (as-received, grinded, grinded and polished and electrolitically polished) and oxidation tests were carried out at 500 and 550ºC from 100 to 2000 hours with two different oxygen concentrations, H2/H2O ratios of 3 and 0.03. In addition, the behaviour of weld joints, T91/T91 and T91/316L, has also been studied under similar conditions. | II 9 |
| 16:45 | Corrosion experiments in flowing LBE at 300 and 450ºC Authors : F.J. Martín-Muñoz*, L. Soler-Crespo, and D. Gómez-Briceño Structural Materials Division, CIEMAT, Building 30 Avda. Complutense 22, Madrid 28040, Spain Tel.: +34 913466631; Fax: +34 913466661 *e-mail: fco.javier@ciemat.es Resume : Lead bismuth eutectic (LBE) has been proposed as a candidate for coolant and/or as spallation target material of Accelerator Driven Systems (ADS) and as coolant for a Generation IV reactor, the Lead-cooled Fast Reactor (LFR). However, one of the critical issues in the feasibility of these systems is the compatibility of steels with the LBE. Therefore, it is important to know the behaviour of steels under conditions similar to which can be expected during the course of reactor operation. Corrosion tests under flowing conditions, using a liquid metal loop, may provide useful results.
A forced convection loop, LINCE, was designed for long-term corrosion tests in LBE at CIEMAT. The volume of LBE is 300 litres and the flow rate approximately 1 m/s. An oxygen control system (OCS) has been implemented in the loop.
The corrosion behaviour of AISI 316L and T91 steels was investigated in flowing lead-bismuth eutectic (LBE) at 300ºC and 450 ºC for 2,000, 5,000 h and 10,000 h. At 300ºC, the results showed a good response, with no weight loss detected in any of the materials after exposure to the flowing LBE up to 10,000 h. A similar behaviour was observed for the probes tested at 450ºC during 2,000 and 10,000 h. However, specimens extracted at intermediate time (5,000 h) showed an anomalous behaviour, with important weight loss. These specimens were placed at the bottom of the hot test section, and this position probably made them to suffer an accused process of cavitation-erosion. | II 10 |
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