Maria Samaras Paul Scherrer Institute
Chu Chun Fu CEA-Saclay
Thak Sang Byun Oak Ridge National Laboratory
Marius Stan Argonne National Laboratory
Toru Ogawa Japan Atomic Energy Agency (JAEA)
Q1: Structural Materials - Steels and Alloys
Monday PM, November 29, 2010
Room 208 (Hynes)
9:30 AM - **Q1.1
WITHDRAWN 12/27/10 Fe-Cr Alloys and Moessbauer Spectroscopy.
Stanislaw Dubiel 1 Show Abstract
1 Physics and Computer Science, AGH University of Science and Technology, Krakow Poland
Moessbauer Spectroscopy (MS) at 57Fe nuclei has proved to be a powerful tool in investigations of Fe-containing materials, among which Fe-Cr alloys, a basic ingredient of stainless steels, play a special role. Thanks both to its high resolution (full line width at half maximum about 5 neV) and to a sensitivity of spectral parameters (mostly that of the hyperfine field) to a presence of Cr atoms in the iron matrix, many questions pertinent to the phase diagram of the Fe-Cr alloy system and related phenomena can be readily studied by means of MS.In particular, the following issues will be addressed in this contribution:(a) short-range ordering,(b) changes in the electronic structure induced by Cr atoms,(c) phase separation into Fe-rich (alpha) and Cr-rich (alpha’) phases,(d) kinetics of the sigma-phase formation,(e) determination of the Curie temperature, and(f) determination of the Debye temperature
10:00 AM - Q1.2
Measurement of Nano-hardness Distribution and Its Application to Embrittlement Measurement of Reactor Pressure Vessel Materials.
Akiyoshi Nomoto 1 , Kenji Nishida 1 , Kenji Dohi 1 , Naoki Soneda 1 Show Abstract
1 , Central Research Institute of Electric Power Industry, Tokyo Japan
Hardness measurement is a simple and useful method to evaluate the irradiation embrittlement of reactor pressure vessel (RPV) materials caused by hardening. However, scatter in data is often a problem to detect slight differences in the amount of embrittlement. This scatter will, in some part, come from the metallurgical inhomogeniety of the materials due to such microstructures like carbides and grain boundaries. In this study, we applied a nano-hardness measurement to evaluate the embrittlement of neutron-irradiated RPV materials. About 4,000 indentations with a 3g force were made within a small area of 2mm x 2mm in order to obtain hardness distribution in the measured area. The measured hardness values distribute in a relatively wide range, but the histogram of the hardness shows a smooth distribution with a clear major peak. Hardness values become higher in the area where carbides are identified with high number density. The special distribution of hardness is also found to correspond very well with the solute atom distributions such as Ni and Mn measured by electron probe micro analyzer. Application of this nano-indentation technique to the detection of hardness increase in RPV steels will also be presented.
10:15 AM - Q1.3
Radiation-induced Ni/Si-rich Precipitates in Austenitic and F-M steels.
Zhijie Jiao 1 , Gary Was 1 Show Abstract
1 Nuclear Engineering and Radiological Sciences, University of Michigan, Ann Arbor, Michigan, United States
Radiation-induced Ni/Si-rich precipitates may cause potential embrittlement in both austenitic and ferritic-martensitic (F-M) steels. However, the nature of the precipitates (including the size, density, composition as well as the nucleation sites) was rarely reported in literature. In this study, 304 stainless steels, CP304 and HP304+Si, and F-M steels, T91, HCM12A and HT-9, were irradiated to 5 dpa at 360°C and 7 dpa at 400°C, respectively, using 2 MeV protons. The Ni/Si-rich precipitates were then examined using atom probe tomography. The results have shown that dislocations are preferable nucleation sites for the Ni/Si-rich precipitates in both austenitic and F-M steels. Manganese, which is a major element of the Ni/Si-rich precipitates in F-M steels, remains as a minor/trace element in the precipitates in the austenitic steels. The common features as well as the differences of the Ni/Si-rich precipitates in austenitic and F-M steels will be presented and discussed.
10:30 AM - Q1.4
Life Limiting Irradiation Damage Phenomena in Reduced Activation Ferritic/Martensitic Steels for Fusion.
Richard Kurtz 1 , G. Robert Odette 2 , Takuya Yamamoto 2 , Yong Dai 3 Show Abstract
1 , Pacific Northwest National Laboratory, Richland, Washington, United States, 2 , University of California, Santa Barbara, Santa Barbara, California, United States, 3 , Paul Scherrer Institut, Villigen Switzerland
The major life limiting irradiation induced degradation phenomena in reduced activation ferritic/martensitic steels for fusion energy systems are reviewed. These phenomena include 1) loss of fracture toughness and ductility at irradiation temperatures below about 400°C, 2) void swelling and irradiation creep at intermediate irradiation temperatures, and 3) thermal creep and helium embrittlement at irradiation temperatures greater than about 600°C. Corrosion, creep-fatigue and related phenomena are also important over a wide range of temperatures. In combination, these damage mechanisms establish neutron fluence, temperature, and stress limits for specific structural applications.In this presentation three issues are considered in some detail. A quantitative model is used to assess the magnitude and implications of irradiation-induced hardening and associated fracture toughness transition temperatures shifts. Emphasis is given to the synergistic effects of high levels of helium produced in the fusion nuclear environment. The potential effects of high helium levels on void swelling are also examined based on a combination of rate-theory models and experiments, including in situ helium injection studies in mixed spectrum reactors. Dimensional stability limits are also assessed including the contributions from both irradiation and thermal creep. The possible limiting influence of high-temperature helium embrittlement is described. The potential for dispersion strengthened ferritic alloys to be more radiation tolerant and expand these limits is briefly discussed.
10:45 AM - Q1.5
Materials Properties of Irradiated Cladding and Duct Reactor Materials.
Tarik Saleh 1 , Stuart Maloy 1 , Mychailo Toloczko 2 , James Cole 3 , T. Byun 4 , Tobias Romero 1 , Joris Van Den Bosch 1 Show Abstract
1 , Los Alamos National Laboratory, Los Alamos, New Mexico, United States, 2 , Pacific Northwest National Laboratory, Richland , Washington, United States, 3 , Idaho National Laboratory, Idaho Falls, Idaho, United States, 4 , Oak Ridge National Laboratory, Oak Ridge, Tennessee, United States
Current plans for burning minor actinides in a transmutation fuel in fast reactors require core materials (cladding and duct) that can withstand very high doses (>300 dpa) while in contact with both coolant and fuel. Research into the mechanical and corrosion behavior of these materials under reactor conditions is conducted under the Fuel Cycle Research and Development program. In this reactor environment the materials would be exposed to radiation effects that promote low temperature embrittlement, high temperature helium embrittlement, swelling, accelerated creep, corrosion with the coolant, and chemical interaction with the fuel (FCCI). Research addressing these issues includes determining radiation effects in ferritic/martensitic steels at doses up to 200 dpa, minimizing FCCI, and developing advanced alloys with improved irradiation resistance. Qualifying materials for use in reactors with fluences greater than 200 dpa requires development of advanced alloys and irradiations in fast reactors to test these alloys. Specimens of ferritic/martensitic alloys (T91/HT-9) previously irradiated in the FFTF reactor to doses of up to 210 dpa at a temperature range of 350-700°C are presently being tested. This includes analysis of a duct made of HT-9 after irradiation to a total dose of 155 dpa at temperatures from 370 to 510°C. Compact tension, charpy and tensile specimens have been machined from this duct and mechanical testing is currently being performed. Complimentary measurements of radiation induced microstructural effects using Small Angle Neutron Scattering and Mossbauer spectroscopy are also underway.
11:00 AM - Q1.6
Atom Probe Characterization of Solute Atom Clustering in Decommissioned Greifswald Unit 4 Weld Metal.
Kenji Nishida 1 , Naoki Soneda 1 , Akiyoshi Nomoto 1 , Kenji Dohi 1 , Frank Bergner 2 , Hans-Werner Viehrig 2 Show Abstract
1 Material Science Reserch Laboratory, Central Reserch Institute of Electric Power Industry, Tokyo Japan, 2 Materials and Components Safety Division, Institute of Safety Research, Forschungszentrum Dresden-Rossendorf, Dresden Germany
Characterization of irradiation-induced microstructural features in reactor pressure vessel (RPV) materials is essential for the understanding of the mechanisms of RPV embrittlement. Particularly, analyses of materials from decommissioned RPVs are extremely important because the simulated irradiations in material testing reactors can never be the same as that of actual commercial reactors. In this study, we performed microstructural characterization of the weld metals cut from the decommissioned Greifswald Unit 4 RPV. Atom probe tomography technique was used to characterize the microstructural features in terms of solute atom clustering and segregation. Materials were cut from two different positions through the wall thickness direction of one weld line. Carbides containing vanadium with a small size of ~10nm in diameter were observed at very low number density. We have also identified the formation of Cu-enriched solute atom clusters which consist of Cu, Ni, Mn, Si, Cr and P. Particularly phosphorus content in clusters is very high in good correspondence with the high bulk P content. In one of the two materials, two kinds of solute atom clusters, i.e. clusters with and without Cu atoms, are observed. Further detained analyses of solute atom clusters will be presented.
Q2: Structural Materials - ODS Alloys I
Monday PM, November 29, 2010
Room 208 (Hynes)
11:30 AM - **Q2.1
Hydrostatic SPD Processing of an ODS Ferritic Steel for Fusion Application.
Malgorzata Lewandowska 1 , Zbigniew Oksiuta 2 , Krzysztof Kurzydlowski 1 Show Abstract
1 Faculty of Materials Science and Engineering, Warsaw University of Technology, Warsaw Poland, 2 Mechanical Department, Bialystok Technical University, Bialystok Poland
Low activation ferritic steels are candidate materials for structural parts of fusion reactors. Such parts are expected to operate at the temperatures reaching 750o C under considerable neutron irradiation. At the same time, the material used for their fabrication should have a high toughness at room temperature. In order to assure such challenging properties, ODS ferritic steels are currently developed within the EUROATOM programme with one of the fabrication routes being based on the powder metallurgy, PM. However, one of the challenges in fabrication via PM route is relatively high porosity of PM ODS ferritic steels and insufficient impact properties.In order to eliminate these drawbacks, various thermo-mechanical treatments have been proposed. Recently, severe plastic deformation, SPD, methods, such as ECAP and HPT, have successfully been employed for nano-refinement of a wide range of metals and alloys. Usually these methods are implemented in ambient conditions, due to the difficulty in maintaining either sub-zero or elevated temperature throughout the process, which last up to a few minutes and in the extreme cases up to a few hours. From that point of view, hydrostatic extrusion, a new processing technology applied to ODS steels, offers considerable advantage as this process is very fast, with the strain rates reaching 102 per second. It should be noted also that such fast deformation is accompanied by adiabatic heating of the billet. This heating is detrimental if low melting point metals are processed and turn out to be useful in the processing of high to deform alloys developed for high temperature/high thermal load applications.The current paper reports results of the study aiming at improving properties and performance of PM ODS ferritic steel by post-sintering processing via hydrostatic extrusion. Earlier results showed that ferritic-martensitic low activation steel undergo nano-refinement as a result of hydrostatic extrusion. Here, we demonstrate that in the case of ODS steels it profoundly reduces residual porosity. It also improves homogeneity of the grain size. On the other hand it does not reduce the size of primary oxides, which affect low temperature toughness of the steel. The results obtained in this study show, that hydrostatic extrusion, or its equivalent, can be used to improve structure/properties of ODS ferritic steels. However, in order to take a full advantage of this processing methods, metallurgical quality of PM billets must be improved.
12:00 PM - Q2.2
Evaluation of Nano-features in Rapidly Solidified Fe14Cr3W0.4Ti0.2Y Nano-structured Ferritic Alloy.
Nicholas Cunningham 1 , A. Etienne 1 , G. Odette 1 , E. Stergar 1 2 , B. Wirth 2 Show Abstract
1 Mechanical Engineering, UC Santa Barbara, Santa Barbara, California, United States, 2 , UC Berkeley, Berkeley, CA, California, United States
Nanostructured ferritic alloys (NFAs) have the potential to make transformational contributions to developing advanced sources of fission and fusion energy. NFAs are Fe-Cr based ferritic stainless steels that contain an ultrahigh density of Y-Ti-O nanofeatures (NFs). The NFs provide for both outstanding high temperature properties and remarkable tolerance to irradiation damage, including mitigation of the degrading effects of transmutation product helium. The results of a detailed study of the evolution, structure and composition of Y-Ti-O NFs in NFA powders and consolidated alloys are described. This work is a small and focused a part of a larger collaboration to produce a large best practice heat NFA. The powders were rapidly solidified by ATI Powder Metals from a melt containing Fe-14%Cr, 3%W, 0.4%Ti and 0.2%Y by gas atomization in Ar, Ar/O, and He atmospheres. Note that this represents a different processing path, since NFA powders are typically processed by mechanical alloying of metallic powders with Y2O3 by ball milling. Proper milling effectively dissolves the Ti, Y and O solutes that then precipitate as NFs during hot consolidation. Electron probe microanalysis (EPMA), atom probe tomography (APT) and small angle neutron scattering (SANS) were used to characterize the powders in the as-atomized, ball milled and ball milled and annealed conditions. The major effect of the quenching gas environment was the O content, which was 0.15 wt% for the Ar/O case and minimized for He and Ar gas quenching. EPMA showed the Y is heterogeneously distributed and phase separated in the as atomized powders. However, attritor milling for 20 to 40 h mixes the Y and adds a significant quantity of O. Effective mixing does not occur at lower milling times. Subsequent powder annealing treatments, typically at 1150°C, result in the precipitation of a high density of NFs. All the annealed powder variants show a bimodal grains size distributions but APT shows presence of NFs in both large and small grains. The NFs in consolidated alloys are also described.
12:15 PM - Q2.3
Cavity Evolution in In-situ Helium Implantation Studies of Nanostructured Ferritic Alloys and Tempered Martensitic Steels.
G. Robert Odette 1 2 , Takuya Yamamoto 1 , Yuan Wu 2 , Dan Edwards 3 , Richard Kurtz 3 Show Abstract
1 Mechanical Engineering, UC Santa Barbara, Santa Barbara, California, United States, 2 Materials Department, UC Santa Barbara, Santa Barbara, California, United States, 3 Materials Department, Pacific Northwest National Lab, Richland, Washington, United States
In-situ He implantation in mixed spectrum fission reactor irradiations provides an attractive approach to assessing the effects of He-dpa synergisms at fusion relevant dpa rates. The idea is to use Ni (or B or Li)-bearing implanter layers to inject high-energy α-particles into an adjacent material that is simultaneously undergoing fast neutron induced displacement damage. For example, 4.8 MeV α-particles are produced by two-step 58Ni(nth,γ)59Ni (nth,α) thermal neutron (nth) reactions. A series of in-situ He implantation irradiation experiments have been carried out in the mixed (fast-thermal) spectrum High Flux Isotope Reactor (HFIR). Micron scale NiAl injector coatings were used to uniformly implant α-particles to a depth of ≈ 5 to 8 μm in transmission electron microscopy (TEM) discs for a large matrix of alloys irradiated over a wide range of temperatures and dpa at controlled He/dpa ratios ranging from << 1 to 40 appm/dpa. These experiments have previously shown the helium is trapped in ultra-small bubbles at the nanofeatures (NFs) in nanostructured ferritic alloys (NFA) at 500°C to 9 dpa and 380 appm He. In contrast a bimodal size distribution of somewhat larger bubbles and even larger voids form in 9Cr tempered martensitic steels (TMS) for identical irradiation conditions. In this talk we describe and compare the cavity structures in a NFA and TMS following HFIR irradiations from 300 to 500°C to ≈ 27 dpa and 1100 appm He.
12:30 PM - Q2.4
High Temperature Mechanical Behaviors of Nanostructured Ferritic Alloy.
Thak Sang Byun 1 , Jeoung Han Kim 1 , David Hoelzer 1 Show Abstract
1 , Oak Ridge National Laboratory, Oak Ridge, Tennessee, United States
The nanostructured ferritic alloys (NFAs) are mechanically-alloyed composite materials of fine-grained ferritic steel matrix and high density nanoclusters. The NFAs have been considered as primary candidates for the core materials of advanced nuclear reactors because of their outstanding strength at high temperatures and high resistance to radiation-induced degradation. In this study the high temperature deformation and fracture behaviors were characterized for one of the latest NFAs: the 14%Cr steel-based 14YWT (SM10 heat) with the composition of Fe-14Cr-3W-0.4Ti (in wt.%) developed at Oak Ridge National Laboratory. First, high temperature tensile deformation and fracture behaviors were characterized for the 14YWT alloy. Tensile tests were performed at various temperatures ranging from room temperature to 1000°C in a vacuum condition at a nominal strain rate of 10-3s-1. Comparing with the ferritic/martensitic steels and conventional oxide dispersion strengthened steels, the 14YWT alloy had much higher strength over the test temperature range, but ductility was significantly lower. A detailed fractography was made for the tested tensile specimens to understand the temperature dependence of fracture mechanism. At relatively low temperatures, the fracture surfaces exhibited a quasi-ductile features presented by a mixture of dimples and cleavage facets. For the specimens tested at 600 – 1000 °C, however, the fracture surfaces were fully covered with dimples. Also, the characteristics of the dimples were dependent on test temperature. Second, the high temperature fracture behavior has been investigated using disk compact tension specimens. The fracture toughness of 14YWT alloy was significantly higher than 100 MPa√m at ≤ 200°C but decreased to significantly lower values. This behavior was explained by the fractography results indicating that the nanostructure of 14YWT alloy produced a unique temperature dependence of fracture surface. The formation of microcracks along aggregate boundaries is believed to be the main reason for the low ductility and toughness of 14YWT alloy. Third, the effects of mechanical testing parameters on stress relaxation behavior were analyzed to compare thermomechanical behaviors of the high strength materials. Finally, the mechanical and microstructural data were used to elucidate the deformation and fracture mechanisms and then to suggest a direction for alloy development.
12:45 PM - Q2.5
Corrosion Behaviour of Martensitic/Ferritic Oxide Dispersed Strengthened Steels in Nitric Medium.
Benoit Gwinner 1 , Martin Auroy 1 , Frédéric Miserque 2 , Michel Tabarant 3 Show Abstract
1 DEN, DPC, SCCME, Laboratoire d'Etude de la Corrosion Non Aqueuse, CEA Saclay, F-91191 Gif sur Yvette France, 2 CEA, DEN, DPC, SCCME, Laboratoire d’Etude de la Corrosion Aqueuse, CEA Saclay, F-91191 Gif sur Yvette France, 3 CEA, DEN, DPC, SCP, Laboratoire Réactivité des Surfaces et Interfaces, CEA Saclay, F-91191 Gif sur Yvette France
The martensitic/ferritic oxide dispersed strengthened (ODS) steels are promising solutions for the fuel cladding of the sodium cooled fast nuclear reactors (SFR). Some studies are currently in progress at CEA to elaborate these steels and demonstrate their good properties (mechanical resistance, ageing behaviour …). Among them a dedicated research program has been launched to evaluate the impact of the corrosion of the cladding on the fuel reprocessing PUREX process. The corrosion behaviour of three martensitic/ferritic ODS steels with different chromium contents (respectively 9, 14 and 18%w.) has been then studied in the fuel dissolution medium (hot and concentrated nitric acid). Immersion tests have been carried out in different experimental conditions (temperature, nitric acid concentration). Corrosion kinetics and morphologies have been characterized by mass loss measurement of samples, by elementary analysis of corrosion products (iron, chromium) in solution by ICP-AES, by optical and scanning electron microscopy and by XPS for the chemical analysis of the oxide layer at the surface.The variation of the corrosion rate with the different parameters (chromium content, temperature, nitric acid concentration) has been studied. The higher the chromium content, the temperature and/or the nitric concentration are, the higher the corrosion rate is. The corrosion dependence with these three parameters has been quantitatively established. The chemical composition of the oxide layer seems to be few influenced by the temperature and the nitric acid concentration, but a lot by the chromium content in the steel. The chromium content in the oxide layer is related to the chromium content in the steel with a linear law. The corrosion is correlated to the chemical composition of the oxide layer. It seems that the ratio Cr/(Cr+Fe) in the oxide layer has to reach a limit value to assure ensure an optimal protection of the steel. With these results we are able to quantify the corrosion rate of the ODS steels cladding in simplified conditions of the fuel dissolution. The next stages of this research program will have the objective to work in more representative conditions by taking into account:1) the presence of the dissolved fuel in nitric acid which can increases the corrosion rate,2) the ageing of the cladding in reactor which can decreases the corrosion resistance of the steels.
Q3: Structural Materials - ODS Alloys II
Monday PM, November 29, 2010
Room 208 (Hynes)
2:30 PM - **Q3.1
Super ODS Steels R&D for Cladding Material of GEN IV Systems.
Akihiko Kimura 1 , Ryuta Kasada 1 , Peng Dou 1 , Noriyuki Iwata 1 , Masaki Inoue 2 , Takanari Okuda 3 , Fujio Abe 4 , Soumei Ohnuki 5 , Shigeharu Ukai 5 , Hiroyuki Sano 6 Show Abstract
1 Institute of Advanced Energy, Kyoto University, Uji Japan, 2 , Japan Atomic Energy Agency, Ibaraki Japan, 3 , Kobelco, Hyogo Japan, 4 , NIMS, Ibaraki Japan, 5 , Hokkaido University, Sapporo Japan, 6 , Nagoya University, Nagoya Japan
The development of high performance fuel cladding is essential for the realization of Gen-IV systems. The 9Cr-ODS martensitic steel was developed as the cladding material for sodium-cooled first breeder reactor in Japan, and the steel showed a good performance in sodium, while the corrosion resistance is poor in supercritical water (SCW) and lead-bismuth eutectics (LBE). High-Cr (>13%) ODS steels added with Al showed a drastic improvement in the corrosion resistance in SCW and LBE. High-temperature strength, however, was remarkably reduced by Al addition because of the characteristic changes in the dispersion morphology of oxide particles. Recently, “super ODS steels” have been developed by means of the third element alloying method, that is, a small addition of Zr, which results in the improvement of high-temperature strength with maintaining high-resistance to corrosion in SCW and LBE. High performance of the super ODS steels stems from the fine nano-scaled (Y, Zr)-oxide particle dispersion. Dispersion morphology of the oxide particles, such as size, number density, chemical compositions, coherency between matrix and particles, are characterized by FE-TEM/EDS observations, high temperature XRD measurements and analyses by FE-EPMA and FE-AES. Chemical compositions of the main particles were significantly influenced by addition of small amount of Zr. They were (Y, Zr) oxide particles rather than (Y, Al) oxide particles. The size and the number density of Zr-added steel was reduced and increased. The coherency of the (Zr, Y) oxide particles depended on the size of the particle. It was also confirmed that a number of oxide particles and carbides exited at grain boundaries in the Zr added ODS steel. Characteristic features of the oxide particles for strengthening mechanism will be summarized for the super ODS steels. Present study includes the result of “R&D of corrosion resistant super ODS steel for highly efficient nuclear systems” entrusted to Kyoto University by the Ministry of Education, Culture, Sports, Science and Technology of Japan (MEXT).
3:00 PM - Q3.2
Investigating the Mechanisms of Formation of Nanoscale Oxide Particles in ODS Steels.
Ceri Williams 1 , George Smith 1 , Emmauelle Marquis 1 Show Abstract
1 Department of Materials, University of Oxford, Oxford United Kingdom
Oxide dispersion strengthened steels are currently regarded as the most promising group of materials for structural applications in fusion power plant and the next generation of high temperature fission reactors. They are designed with a reduced activation composition, and are based on either a ferritic or ferritic/martensitic structure to minimise swelling under irradiation. Usually, the base alloy is mechanically alloyed with 0.3-0.5wt.%Y2O3 then subjected to hot isostatic pressing or extrusion to introduce a fine dispersion of nanoscale oxide features. As well as improving the high temperature properties of the material compared to the base alloy, the oxide nano-features act as trapping sites for helium and point defects introduced during irradiation. The size distribution, number density and composition of the oxide nanoclusters are highly dependent on both the composition of the starting powders, and the processing parameters.Understanding how the nanoclusters form and develop is key to refining the mechanical alloying process and optimizing the microstructure. Two ODS alloys, developed by Baluc et. al. (EPFL, Switzerland) were investigated during this study. The material has an Fe-14%Cr-2%W-0.3%Ti base alloy composition which is mechanically alloyed with either 0.3wt% Y2O3 or 0.5wt.% Fe2Y. The oxide nanoclusters have been characterised at various stages in the processing route using atom probe tomography (APT) and transmission electron microscopy (TEM).Using focussed ion beam milling techniques, specimens suitable for APT and TEM were created from the milled powder. APT analysis confirms that Y and O form a supersaturated solid solution in the powder after milling. A series of annealing treatments on the milled powder shows the initial stages of formation of the nanoclusters.When analysing the consolidated material, both starting powders resulted in a dispersion of oxide particles, yet the size distribution and the composition of the particles varied substantially between the two alloys. In the 0.3wt% Y2O3 material, the oxide clusters are ~2nm diameter and have a Ti:Y ratio of approximately 2:1. However, in the 0.5wt.% Fe2Y material, the clusters are ~5nm in diameter and are primarily Y-O based, with very little Ti. This indicates a different particle formation mechanism, and suggests that controlling both the Ti/Y ratio, and the proportion of oxygen are crucial to the formation of the nanoclusters. Through high temperature annealing of the consolidated material, coarsening mechanisms of the nanoclusters is investigated, giving greater understanding of the stability of these features.
3:15 PM - Q3.3
Characterization of Oxide Particles in ODS Austenitic Stainless Steel after Heavy Ion Irradiation up to High Dose.
Hiroshi Oka 1 , Yosuke Yamazaki 1 , Hiroshi Kinoshita 1 , Naoyuki Hashimoto 1 , Somei Ohnuki 1 , Shinichiro Yamashita 2 , Satoshi Ohtsuka 2 Show Abstract
1 Faculty of Engineering, Hokkaido University, Sapporo Japan, 2 O-arai Research and Development Center, JAEA, O-arai, Ibaraki, Japan
The oxide dispersion strengthened (ODS) steels has excellent thermal creep resistance due to a high density of small oxide particles dispersed in the matrix. In our laboratory, ODS austenitic stainless steels based on an advanced SUS316 steel has been developed by mechanically alloying (MA) and hot extrusion with the addition of minor alloying elements. We focused on the stability of oxide particles in ODS austenitic stainless steels (ODS316) irradiated up to high doses.The chemical composition of ODS316 is Fe-16Cr-13Ni-0.35Y2O3-0.1Ti-0.6Hf. The ODS316 has been irradiated with Fe+ ion up to 100 dpa at 500 C using an ion-accelerator operated at 250 keV at Hokkaido University. A TRIM calculation indicated that a damage region was distributed to 100 nm from the surface. Thin foils for transmission electron microscope (TEM) examination were prepared with electro-polishing for un-irradiated side of specimen. HRTEM and EDS were used for characterization of oxide particles in ODS-316, especially precipitation behavior and chemical composition before and after irradiation. Microstructural observation revealed that the ODS316 has grains of 0.5-1 μm in diameter and complex oxides (Y-Ti-Hf-O) distributed in matrix. Complex oxides seemed to be stable under the heavy-ion irradiation in the present condition. On chemical composition of complex oxide particles, the ratio of Ti appeared to be decreased after irradiation. The strain contrast was observed around oxide particles, suggesting coherency of oxide particles. HR-TEM observation revealed that a part of faceted particles have coherency before irradiation. Interface between matrix and oxide particles after irradiation was also investigated.
3:30 PM - Q3.4
On the Formation Mechanism of Oxide Nanoparticles in Fe-Cr MA/ODS Steels.
Luke Hsiung 1 , Michael Fluss 1 , Akihiko Kimura 2 Show Abstract
1 Physical and Life Sciences, Lawrence Livermore National Laboratory, Livermore, California, United States, 2 Institute of Advanced Energy, Kyoto University, Kyoto Japan
Structure of oxide nanoparticles in Fe-16Cr-4.5Al-0.3Ti-2W-0.37Y2O3 (K3), Fe-20Cr-4.5Al-0.34Ti-0.5Y2O3 (MA956), and Fe-14Cr-1.0Ti-0.3Mo-0.25Y2O3 (MA957) ODS steels produced by mechanical alloying (MA) method has been studied using high-resolution transmission electron microscopy (HRTEM) techniques to better understand the formation mechanism of oxide nanoparticles in ODS steels. Partially crystallized nanoparticles with a core/shell structure were frequently observed in all three ODS steels. While the crystalline nanoparticles in K3 and MA956 steels are mainly Y4Al2O9, those in MA957 steel are mainly Y2TiO5 and Y2Ti2O7. HRTEM observations of crystalline nanoparticles larger than ~2 nm and amorphous or disordered cluster-domains smaller than ~2 nm provide us an insight into the formation mechanism of oxide nanoparticle in MA/ODS steels that involves solid-state amorphization and crystallization. This work was performed under the auspices of the U.S. Department of Energy by Lawrence Livermore National Laboratory under Contract DE-AC52-07NA27344.
Q4: Structural Materials - Stainless Steels
Monday PM, November 29, 2010
Room 208 (Hynes)
4:15 PM - **Q4.1
Development of EHP Grade Austenitic Stainless Steels for Nuclear Plants.
Junpei Nakayama 1 , Kiyoshi Kiuchi 1 Show Abstract
1 , KOBE STEEL, LTD., Kobe Japan
The new type austenitic stainless steel so-called EHP (Extra-High-Purity) is developed for improving the resistance against environmental assisted cracking during plant aging. It is corresponding to the highest grade with the major impurities controlled less than 100ppm. It is only achieved with the two-stepped refining process including the electron beam melting for sufficiently eliminating the volatile minor elements in relation to the solidification points and the metallic bonding. The high Cr-Ni corrosion resistant EHP steels with the high austenitic stability are easy to apply without the risk to cracking during melting and welding processes by raising the eutectic point affected with impurities. The resistance to IGC and IGSCC is possible to improve in the wide corrosion potential range expected in the commercial use. The new standards including the steel making process, the welding procedure and the chemical composition range are selected for maintaining the excellent characteristics of EHP steels.
4:45 PM - Q4.2
Crevice Corrosion Behavior of Type 316L Stainless Steel in Gamma-ray Irradiated High-temperature Water.
Yukio Nakahara 1 , Chiaki Kato 1 , Masahiro Yamamoto 1 , Takashi Tsukada 1 , Atsushi Watanabe 2 , Motomasa Fuse 3 Show Abstract
1 Nuclear Science and Engineering Directorate, Japan Atomic Energy Agency, Tokai-mura, Naka-gun, Ibaraki Japan, 2 Energy and Environmental Systems Laboratory, Hitachi, Ltd., Hitachi-shi, Ibaraki Japan, 3 Hitachi Works, Hitachi-GE Nuclear Energy, Ltd., Hitachi-shi, Ibaraki Japan
It is well known that the decomposed compounds of water by irradiation affect the corrosion of structural materials, i.e. stainless steel (SS), in light water reactors. We utilized a gamma-ray irradiated autoclave in order to clarify the irradiation effects on the corrosion of SS in high-temperature water. The corrosion in diffusion-restricted environment was also examined with simulated crevice-shape samples. Test specimens were made of Type 316L SS. The specimens have been immersed in deaerated high purity water of 288 oC with gamma-ray irradiation for 500 hours. The source of the gamma rays was 60Co, The maximum absorbed dose rate in the water was estimated to be about 30 kGy h-1. The shape of the specimens was disk-like, 16 mm in diameter by 0.5 mm in thickness. The surfaces of the specimens were finished with #800 emery paper. The simulated crevice-shape sample was made by fixing two specimens contacting with each other closely. Some specimens were immersed as the crevice-shape samples, and the others were immersed alone for comparison. Characteristics of the surface oxide formed on tested specimens have been analyzed using SEM, laser Raman spectrometer (LRS), and TEM / EDX. The surface of the specimens immersed alone was covered with precipitated particles. The size of the particles was 1 micrometer or less in diameter. The spectrum of LRS indicated that the particles were Fe-Ni spinels (NiFe2O4). Some peaks of α-Fe2O3 were observed in the spectrum. In-depth profiles of the surface taken using TEM / EDX showed that the structure of surface oxide was two layers. The outer layer, which corresponds to the precipitated particles, consisted of Fe-Ni spinels. The inner layer, which is assumed to be the oxide of the original specimen surface, was oxide of Fe, Cr, and Ni. On the surface of the specimens immersed as the crevice-shape samples, precipitated particles were also observed but the surface was not fully covered with the particles. The size of the particles was 1 micrometer or less in diameter. Besides the particles, bigger particles were observed on the surface. The size of the bigger particles was about 5 micrometers in diameter. The LRS spectra indicated that the smaller particles were Fe-Ni spinels and the bigger particles were α-Fe2O3. In-depth profiles of the surface taken using TEM / EDX showed that the smaller particles were Fe-Ni spinels. The in-depth profile and electron diffraction of the bigger particle indicated that it was α-Fe2O3. The profiles also indicated that the original surface was oxidized. The oxide layer contained Fe, Cr, and Ni. The oxide layer of the gamma-ray irradiated specimens contained α-Fe2O3 and bigger particles of α-Fe2O3 were observed in only crevice portion. It indicated that oxidant created by the irradiation changed the corrosion potential to nobler direction and the effect was dominant at the crevice portion where the diffusion of oxidant was restricted.
5:00 PM - Q4.3
In-Situ TEM Study of Radiation Induced Segregation Mitigation by Σ3 Grain Boundaries in Fe and Fe-Ni-Cr.
Greg Vetterick 1 , Christopher Barr 1 , Dan Scotto D'Antuono 1 , Mark Kirk 2 , Mitra Taheri 1 Show Abstract
1 Materials Science and Engineering, Drexel University, Philadelphia, Pennsylvania, United States, 2 , Argonne National Laboratory, Argonne, Illinois, United States
As always, there is a drive for cleaner, safer, and more efficient energy. The current focus on climate change and domestic dependence of energy provides additional motivation to extend the lifetimes of current light water reactors. In LWRs, austenitic steels are subject to irradiation at elevated temperature. This can lead to failure due to radiation damage mechanisms at grain boundaries, such as radiation induced segregation (RIS). It has been shown the lifetime of steel components in nuclear reactors can be increased using grain boundary engineering (GBE). The use of thermomechanical processing to create higher quantities of low-index coincidence site lattice (CSL) grain boundaries improves resistance to creep, stress corrosion cracking, and embrittlement of the material. This improvement in mechanical properties is thought to arise due the mitigation of grain boundary depletion and enrichment; however, the precise mechanism is not well understood. To reliably improve the mechanical properties of austenitic stainless steels for nuclear applications, an improved understanding of damage mechanisms as a function of grain boundary character (GBC) is needed.This study strives to determine how GBC affects defect nucleation, clustering, and solute segregation, in irradiated steels. In particular, we investigate low energy Sigma-3 (Σ3) grain boundaries in simulated LWR conditions. Model Fe and Fe-Ni-Cr-type alloys are compared in as-received and GBE conditions. In the GBE condition, a high fraction of low angle coincident site lattice boundaries (Σ3 CSL) was induced. These samples were irradiated at elevated temperatures and pulled in tension using in-situ transmission electron microscopy. Combined with back scattered diffraction (EBSD) and high resolution, ex-situ TEM with STEM-EDS, some insight of the effect of GBC on RIS can be obtained. The in-situ electron microscopy was accomplished at the Electron Microscopy Center for Materials Research at Argonne National Laboratory, a U.S. Department of Energy Office of Science Laboratory operated under Contract No. DE-AC02-06CH11357 by UChicago Argonne, LLC.
5:15 PM - Q4.4
Ni-Si Phases in Irradiated Austenitic Steels.
Zhijie Jiao 1 , Jan Michalicka 1 , Gary Was 1 , Chen Ling 2 , Anton Van der Ven 2 Show Abstract
1 Nuclear Engineering and Radiological Sciences, University of Michigan, Ann Arbor, Michigan, United States, 2 Materials Science and Engineering, University of Michigan, Ann Arbor, Michigan, United States
Radiation-induced Ni-Si phases (including γ' and G phases) are frequently observed in austenitic steels. However, the formation and evolution of those phases under irradiation are not well understood. In this study, the commercial grade 304 stainless steels were irradiated to 5 dpa at 360°C using 2 MeV protons and up to 100 dpa at 500°C using heavy ions. The nature of the Ni-Si phases was examined using a combination of atom probe tomography and transmission electron microscopy. Preliminary results have shown that most Ni/Si-rich precipitates observed at 5 dpa have yet to reach the stoichiometric compositions of the γ' or the G phases. Stability of the Ni-Si phases will also be evaluated as a function of local alloy concentration, and interstitial and vacancy supersaturations from first principles using density functional theory. The stable Ni-Si phases predicted by the modeling will be compared to the experimental observations.
5:30 PM - Q4.5
In-situ TEM Ion Irradiation of an Advanced Austenitic Stainless Steel.
Meimei Li 1 , Marquis Kirk 1 , P. Baldo 1 , Ken Natesan 1 Show Abstract
1 , Argonne National Lab, Argonne, Illinois, United States
The HT-UPS (high-temperature, ultrafine precipitation-strengthened) steel is a 14Cr-16Ni-2.5Mo-2Mn austenitic stainless steel modified with a combination of V, Nb, Ti, B, P to form stable nano-scale precipitates for high-temperature strengthening. It was developed in the late 1980’s to improve the combination of resistance to radiation-induced swelling and grain boundary helium embrittlement, and resistance to thermal creep-rupture for fusion energy applications. It is currently being evaluated for its potential applications in advanced fast reactors. This paper presents the results of in-situ ion irradiation experiments of the HT-UPS steel within a transmission electron microscope (TEM). The alloy was irradiated at the IVEM-Tandem facility at Argonne National Laboratory with 1 MeV Kr ions at room temperature and 300C to doses up to ~4 dpa. The development of defect structure was followed at successive ion irradiation doses. Quantitative analysis of TEM images was made to determine the number density and size of defect clusters as a function of irradiation dose and temperature. Post-irradiation annealing experiments were performed at incrementally increased temperatures to examine defect structure changes during annealing. The effect of thermal-mechanical treatments on irradiated microstructure was investigated by examining the alloy in two different thermal-mechanical treatment conditions, i.e., solution-annealed and hot-rolled.
Maria Samaras Paul Scherrer Institute
Chu Chun Fu CEA-Saclay
Thak Sang Byun Oak Ridge National Laboratory
Marius Stan Argonne National Laboratory
Toru Ogawa Japan Atomic Energy Agency (JAEA)
Q6: In-Room Poster Session: Nuclear Materials & Fuels
Tuesday AM, November 30, 2010
Room 208 (Hynes)
Q5: Plasma Facing Materials
Tuesday PM, November 30, 2010
Room 208 (Hynes)
9:30 AM - Q5.1
Fracture Behavior of Tungsten Materials and the Impact on the Divertor Design in Nuclear Fusion Power Plants.
Michael Rieth 1 , Andreas Hoffmann 2 , Jens Reiser 1 , David Armstrong 3 Show Abstract
1 Institute for Materials Research I, Karlsruhe Institute of Technology, Karlsruhe Germany, 2 Development Refractory Alloys, PLANSEE Metall GmbH, Reutte Austria, 3 Department of Materials, University of Oxford, Oxford United Kingdom
Present US and European helium cooled DEMO divertor design studies make use of the high temperature strength and good heat conductivity of tungsten. In such outlines, refractory materials are used for structural parts. The most critical issue of tungsten materials in connection with structural applications is the ductile-to-brittle transition (DBT) which is already at rather high temperatures for tungsten materials. Depending on irradiation and dynamic load conditions, the DBT level could be even higher. A systematic study of microstructure and impact bending properties of standard tungsten materials was performed on different tungsten rod, plate, and round blank materials. The results of the tungsten materials look quite different compared to those of typical bcc metals. Only specimens of pure tungsten show a clear upper shelf area, starting at about 800 °C. All other rod materials don’t show pure ductile fracture within the whole test temperature range. On the other hand, all tested materials tend to brittle fracture at temperatures below 500 °C. But above that temperature, the specimens show delamination fractures which propagate along the rod axis, that is, parallel to the specimen’s long side and perpendicular to the notch. Compared to the rods, all plate materials show even worse properties: (1) the energies are lower by more than 50 %, and (2) the fracture is dominated by delamination within the whole temperature range. In summary, there are three types of fracture behaviour (brittle, ductile, and delamination) which is closer correlated to the materials microstructure than to the materials composition. This paper reviews the results and other relevant properties of tungsten materials with respect to application for structural divertor parts. Drawbacks and possible alternatives are discussed.
9:45 AM - Q5.2
Nanoindentation and Micro-mechanical Testing of Self ion Implanted Tungsten.
David Armstrong 1 , Xiaoou Yi 1 , Angus Wilkinson 1 , Steve Roberts 1 Show Abstract
1 Department of Materials, University of Oxford, Oxford United Kingdom
There is a great deal of interest in using tungsten alloys as critical plasma facing components in the divertor. For this to occur the mechanical properties, and how they change under the extreme conditions faced in the power plant must be understood. Of particular importance are the effects of neutron damage on mechanical properties.Neutron irradiation in a fusion reactor will cause a large amount of damage to the crystal structure and also transmutation of W to Re and Os, producing He, and hence degrade the mechanical properties of the material. Low-energy neutron irradiation of candidate materials is possible; however it can take several years to create significant levels of damage. Hot cells and remote handling facilities are then required for testing the resulting active samples. Ion-implantation can be used to mimic the displacement damage produced by neutrons. However the depth over which the damage occurs is only a few hundred nanometres to a few microns in depth. Thus standard mechanical testing techniques are not suitable for measuring changes in mechanical properties due to implantation. Specimens of pure tungsten, W-5wt%Ta, W-1wt%Ta and W-5wt%Re have been irradiated using an ion-beam facility at the University of Surrey, UK. 2 MeV tungsten ions were used, at 500oC, producing a damaged layer of depth ~200nm with doses aimed at generating damage levels of 0.5 displacement per atom (dpa),1dpa, 5dpa,15dpa and 50dpa. The damage has been studied using TEM and is mostly <111> type loops with a density of 5x1017/m2 and average size of 4-6nm at 15dpa. High Resolution-EBSD has been used to measure the elastic strains found in the implanted layer as compared to unimplanted regions.Nanoindentation has been performed using an MTS Nano XP. An increase of ≈5% to 30% in hardness in the implanted layer is seen as the dose is increased. The depth at which this occurs over is in agreement with the depth of 200nm predicted by SRIM. The increase in hardness also varies with composition; the W5Ta alloy has a much larger increase in hardness, than the W5Re alloy which in turn has a larger increase than the pure W. As well as the increase in hardness an increase in elastic modulus is also seen. This magnitude of this is independent of dose and composition and is ≈15% of the unimplanted value. Due to the poorly defined stress states around the indenter extracting parameters such as yield stress and work hardening is difficult from indentation experiments. Recently novel micro-mechanical-testing techniques have been developed based on the imaging and loading of FIB machined micro-cantilevers using a nanoindenter. Work is now ongoing to reduce the size of micro-cantilevers to allow the mechanical properties of the very shallow implanted layers in tungsten alloys to be measured. This will allow yield stress and work hardening of the ion implanted damaged layer to be measured directly without influence of the underlying material.
10:00 AM - Q5.3
Precipitation in Irradiated W Alloys.
Emmanuelle Marquis 1 2 , Colin English 1 , Samuel Humphry-Baker 1 Show Abstract
1 Department of Materials, University of Oxford, Oxford United Kingdom, 2 Materials Science and Engineering, University of Michigan, Ann Arbor, Michigan, United States
Tungsten is being considered for plasma facing and diverter components in future fusion reactors. A major issue for the use of tungsten is the transmutation of W into Re and Os under neutron irradiation, and current neutronic calculations (Cottrell, J. Nucl. Mater. 2004) predict that pure tungsten would transform into W-13%Re-12%Os after just five years of service. As demonstrated in past studies by Seidman et al. (acta Metall. 1984) and Williams et al. (Metall. Trans. A 1983), irradiated W-Re alloys exhibit precipitation of Re-rich phases even in undersaturated solid-solutions. The precipitation mechanisms are unclear and understanding the kinetics of precipitation is a necessary step towards establishing predictive tools for the structural and mechanical integrity of the reactor components during service, thereby ensuring reactor safety and optimizing reactor efficiency. We report on a systematic series of atomic scale investigations on W alloys implanted using 2MeV W+ ions. The formation of solute-rich clusters and the kinetics of precipitation will be discussed.
10:15 AM - Q5.4
Simulation of He Defect Properties, Diffusion and Bubble Formation in W.
Niklas Juslin 1 , Brian Wirth 1 Show Abstract
1 Nuclear Engineering, UC Berkeley, Berkeley, California, United States
Helium will be present in fusion reactor materials, such as the proposed tungsten divertor in ITER, due to He bombardment from the plasma, as well as transmutation reactions caused by high energy neutron interactions. Helium is known to form bubbles and change the material properties of tungsten, e.g. swelling and ductile to brittle transformation temperature. Furthermore, an interesting surface modification of tungsten exposed to low-energy He plasmas at high temperatures has been observed in experiments, but the formation of tendrils of W-He, known as "nano-fuzz", is still not understood.Molecular dynamics (MD) and Monte Carlo (MC) simulations are important tools to study the behavior of helium in tungsten on an atomistic level, which is not easily achieved in experiments, and on longer length and time scales than those available in ab initio calculations. In order to identify the correct mechanisms and trustworthy qualitative and quantitative results, the inter-atomic potentials used in the simulations are of utmost importance.We present new potentials for the simulation of He in W. The potential for W is a modification of the Ackland-Thetford potential with improved interstitial formation energies and threshold displacement energies. For W-He, existing potentials do not describe helium defects in accordance with literature density functional theory (DFT) data and a new repulsive pair potential has been developed with emphasis on reproducing the formation energies of substitutional and interstitial He and the binding of two He atoms in the tungsten matrix.The W-He potential has been tested by simulating the diffusion of He and comparing the formation and binding of small He-vacancy clusters to DFT data. In order to understand He bubble formation in tungsten, we have studied He and vacancy binding to He-vacancy clusters of different sizes, and the stability and mobility of clusters small enough to be mobile on the time scale available in MD simulations. These simulations help with the description of He bubble formation and provide important parameters for future MC simulations.
10:30 AM - Q5.5
Investigating Diamond as a Plasma Facing Material for Fusion Power.
Alastair Dunn 1 , Dorothy Duffy 1 , Philip John 2 , Samuele Porro 2 , Marshall Stoneham 1 , Greg de Temmerman 3 , John Wilson 2 Show Abstract
1 Physics and the London Centre for Nanotechnology, University College London, London United Kingdom, 2 School of Engineering and Physical Sciences, Heriot-Watt University, Edinburgh United Kingdom, 3 , FOM Institute for Plasma Physics Rijnhuizen, Nieuwegein Netherlands
The intense heat and particle fluxes predicted for future fusion reactors make the selection of materials for the plasma facing components a challenging problem. This is particularly true for the divertor, the region at the base of the tokamak where the heat and the fusion products are extracted. Carbon, in the form of graphite or fibre reinforced composites, is favoured as a plasma facing material in plasma experiments, mainly because of its high thermal conductivity and low Z, which means that any material entering the plasma will have minimal radiative cooling. However the high reactivity of C with hydrogen isotopes has caused concern about the tritium inventory for deuterium – tritium (D-T) plasmas. The low energy hydrogen atoms interact with surface C atoms to produce volatile hydrocarbon species that redeposit elsewhere in the vessel. The tritium transported and deposited in such reactions would be difficult to recover, which would present a safety hazard. Diamond has been proposed as a potential plasma facing material because of its excellent thermal and mechanical properties and its higher resistance to hydrogen erosion than graphitic materials. In this paper we will present results from a combined experimental/modelling project designed to investigate the potential for CVD diamond as a plasma facing material. We will report results from trials of microcrystalline and nanocrystalline diamond, deposited by hot filament chemical vapour deposition on molybdenum, graphite and silicon, in a range of tokamaks (MAST (UK), DIII-D (USA) and TEXTOR (Germany)) and in the Pilot-PSI (Netherlands) linear plasma source. The samples were characterised before and after exposure by SEM & TEM, Raman spectroscopy, XPS, NRA, and mechanical tests. In none of these tests did we observe catastrophic failure or delamination of diamond coatings. We will describe the modest changes to the diamond surface from plasma exposure, and present erosion rates and deuterium retention amounts. The experiments were supported by modelling studies in which molecular dynamics was used to bombard diamond surfaces with 15 eV tritium atoms, in order to investigate tritium retention and chemical erosion rates as a function of surface temperature. As with the experimental investigations, the results from the modelling studies suggest that diamond may have superior properties to graphitic carbon. The tritium retention was limited to the first few layers near the surface. The surface layers became disordered as the C atoms were etched by the tritium, but the diamond structure persisted and there was no evidence that amorphous carbon was formed below the surface layers. At temperatures above 900 K the diamond structure transformed to graphite and this resulted in deeper tritium penetration than that observed in the substrates in which the diamond structure survived.
Q6: In-Room Poster Session: Nuclear Materials & Fuels
Tuesday PM, November 30, 2010
Room 208 (Hynes)
11:30 AM - Q6.1
Two Advance Structural Alloys for New Generation Nuclear Systems.
Mikhail Sokolov 1 , Yukinori Yamamoto 1 , Lizhen Tan 1 Show Abstract
1 , ORNL, Oak Ridge, Tennessee, United States
Fracture toughness and tensile properties of two advanced structural alloys were characterized in the wide temperature range from room temperature to 650C. These two alloys are the candidates of high-temperature structural materials in the next generation nuclear systems. The HT-UPS (High Temperature Ultrafine Precipitate Strengthened) austenitic stainless steel with Fe-14Cr-16Ni (wt%) base composition exhibits superior creep strength compared to the advance austenitic stainless. A columnar shape cast ingot with 4” in diameter and 5” length was hot-forged and hot-rolled at 1200C, which resulted in a plate sample with the size of 0.8” x 0.5” x 16”. The plate showed uniform solutionized microstructure with an average grain size of around 100 um, together with a lot of TiN dispersions with less than 5 um size which was introduced during solidification process. As result, tensile tests did not reveal any significant differences for specimens in the rolling and transverse directions. Other candidate alloy in this study was ferritic-martensitic 9Cr-1.8WNbV steel NF616. Both alloys exhibited considerable fracture toughness properties up to 650C. In addition, thermomechanical treatment is being developed on the as-received NF616. Preliminary results of Vickers microhardness and tensile tests showed enhancement in microhardness and both strength and ductility compared to the as-received NF616. Microstructural characterization is being performed to explain the properties enhancement induced by the thermomechanical treatment.
11:30 AM - Q6.10
Magnetic Tight Binding Simulations of Defects in Iron.
Preetma Soin 1 2 , Andrew Horsfield 1 , Duc Nguyen-Manh 2 , Adrian Sutton 3 Show Abstract
1 Materials, Imperial College, London United Kingdom, 2 , Culham Centre for Fusion Energy, Abingdon United Kingdom, 3 Physics, Imperial College, London United Kingdom
Iron is a key element in the ferritic steel used in hydrogen fusion power plants. Since we are unable to observe directly the point defects generated by irradiationwe use atomistic simulations. As empirical potentials do do not give a proper description of the magnetic properties and DFT becomes too computationally expensive for large unit cells, we use a magnetic tight binding scheme.We extended an existing Stoner model that employed local charge neutrality to include both onsite and intersite Coulomb interactions. A new charge mixing scheme has been devised to stabilise and accelerate the convergence to self-consistency. We have carried out very well converged simulations using our parallelised code, PLATO. So far we have investigated vacancies and interstitials in ferritic iron. We observe the dependence of the local magnetic moment and charge transfer on the structure of the defects. We will discuss future applications to line and planar defects.
11:30 AM - Q6.11
The Effect of Cr Concentration on Defect Energies in FeCr Alloys.
Emma del Rio 1 , Maria Caturla 2 , Jesus Sampedro 1 , Manuel Perlado 1 Show Abstract
1 , Instituto Fusion Nuclear - Universidad Politecnica de Madrid, Madrid Spain, 2 Departamento de Fisica Aplicada, Universidad de Alicante, Alicante Spain
High-Cr ferritic/martensitic steels are leading candidates for key components in most future nuclear applications. The good resistance against corrosion and the low damage accumulation and swelling are the main reasons for this. Nevertheless, these alloys present problems of embrittlement. The presence of defects created by the irradiation could be responsible for this effect, since they are obstacles for the motion of dislocations. Therefore, the mechanical response of these materials will depend on the type of defects created during irradiation. Transmission electron microscopy (TEM) experiments have shown that the concentration and type of defects observed depend on Cr concentration, among other factors. It is well known that the addition of Cr to the steels improve their properties against radiation damage. But the behaviour of FeCr alloys present a non-monotonic trend of radiation hardening, embrittlement or swelling vs Cr concentration. Understanding this effect is very important for the development of reliable models beyond empirical correlations.The main objective of this work is to understand, at a fundamental level, the stability of different types of defects as a function of Cr content. In order to do this study molecular dynamics and static calculations are performed using the two available interatomic potentials for FeCr alloys (A. Caro et al. and P. Olsson et al.). The stability of small clusters (up to size 5) of vacancies and self-interstitials for different Cr content is analyzed. On the other hand, self-interstitial clusters of sizes in the range that can be observed experimentally (~2nm) are studied. In this case two types of clusters are considered <111> and <100> loops, since those are the ones observed by TEM. Both pure Fe loops and mixed loops (containing Cr atoms) are studied.
11:30 AM - Q6.12
Atomistic Modeling Study of Point Defect Interactions with Coherent, Nanoscale Precipitates.
Tuan Hoang 1 , Hyon-Jee Lee 1 , Brian Wirth 1 2 Show Abstract
1 Nuclear Engineering, University of California, Berkeley, Berkeley, California, United States, 2 Nuclear Engineering, The University of Tennessee-Knoxville , Knoxville, Tennessee, United States
Materials used in extremely hostile environment such as nuclear reactors are subject to a high flux of neutron irradiation, and thus vast concentrations of vacancy and interstitial point defects are produced because of collisions of energetic neutrons with host lattice atoms. The fate of these defects depends on various reaction mechanisms, which occur immediately following the displacement cascade evolution and during the longer-time kinetically dominated evolution. In this work, we use atomistic modeling to evaluate the effect of nanometer-sized, coherent precipitates on the ability to attract and annihilate point defects. In particular, we have evaluated the effect of size and elastic modulus mismatch on the interaction with point defects using molecular dynamics. Sutton-Chen N-body potentials have been used to describe the many-body interactions between atoms in generic alloy systems, and provide the ability to systematically evaluate the effect of lattice parameter or elastic modulus changes on the interaction range of nanoscale precipitates and point defects. The stress and strain field generated by the presence of the nanoscale precipitates is also studied by means of continuum mechanics and results are compared with those obtained from atomistic simulation. This study provides insights into the underlying process of microstructural evolution in nuclear materials and into the important characteristics for the manufacture of advanced, radiation resistant structural materials used in high radiation environment such as next generation fission and fusion reactors.
11:30 AM - Q6.14
Low Temperature Synthesis and Sintering of d-U, d-U/La Alloy and d-UO2 Nanoparticles.
Benjamin Jacobs 1 , David Robinson 1 , Jianyu Huang 2 , Summer Ferreira 2 , Paula Provencio 2 , Tina Nenoff 2 Show Abstract
1 , Sandia National Labs, Livermore, California, United States, 2 , Sandia National Labs, Albuquerque, New Mexico, United States
We report on the novel room temperature method of synthesizing advanced nuclear fuels; a method that virtually eliminates any volatility of components. This process uses radiolysis to form stable nanoparticle (NP) nuclear transuranic (TRU) fuel surrogates and in-situ heated stage TEM to sinter the NPs. The radiolysis is performed at Sandia’s Gamma Irradiation Facility (GIF) 60Co source (3x10^6 rad/hr). Using this method, sufficient quantities of fuels for research purposes can be produced for accelerated advanced nuclear fuel development. We are focused on both metallic and oxide alloy nanoparticles of varying compositions, in particular d-U, d-U/La alloys and d-UO2 NPs. We present detailed descriptions of the synthesis procedures, the characterization of the NPs, the sintering of the NPs, and their stability with temperature. We have employed UV-vis, HRTEM, HAADF-STEM imaging, single particle EDX and EFTEM mapping characterization techniques to confirm the composition and alloying of these NPs.Sandia is a multiprogram laboratory operated by Sandia Corporation, Lockheed Martin Company, for US DOE’s NNSA, Contract DE-AC04-94-AL85000.
11:30 AM - Q6.15
Grain Boundary Chemistry in Irradiated Fe-Cr Alloys.
Rong Hu 1 , Emmanuelle Marquis 1 2 Show Abstract
1 Department of Materials, University of Oxford, Oxford United Kingdom, 2 Materials Science and Engineering, University of Michigan, Ann Arbor, Michigan, United States
Ferritic Fe-Cr alloys are the base for structural steels currently considered for Gen IV fission nuclear plants and future fusion reactors. Among the outstanding issues related to the use of these alloys under irradiation are α’ precipitation (and the uncertainty of the position of the solvus line at low temperatures) and radiation induced segregation (RIS) or depletion (RID) of grain boundaries. These phenomena could indeed significantly impact strength, fracture and corrosion resistance of these alloys. Past observations have reported irradiation-induced depletion or segregation without any clear correlation to irradiation dose, dose rate, alloy concentration or temperature (Lu et al. Scripta Mater. 2008). Moreover, common impurities such as carbon are expected to affect the integrity of grain boundaries and modify the extent of RIS/RID phenomena. The experimental approach based on atomic scale characterization techniques for investigating RIS and RID in these alloys as well as the role of carbon will be described and the effect of irradiation discussed as function of carbon content and grain boundary orientation.
11:30 AM - Q6.3
Kinetic Lattice Monte Carlo Simulations of Radiation Induced Segregation of Chromium in Ferritic-martensitic Steels.
Brian Frisbie 1 , Brian Wirth 1 Show Abstract
1 Nuclear Engineering, University of California, Berkeley, Concord, California, United States
Ferritic-martensitic steels with 8-12% Cr are leading candidates for advanced nuclear applications because of their superb thermal properties and swelling resistance. Irradiation of these steels introduces microstructural and microchemical changes in the material that can lead to degradation of their structural properties through processes such as radiation-induced embrittlement. In this work, we present a kinetic lattice Monte Carlo model intended to simulate the segregation behavior of Cr to grain boundaries in these ferritic-martensitic steels while under irradiation. The simulations confirm that the enrichment or depletion of Cr at grain boundaries is dependent on its diffusivity relative to that of Fe. The spatial distribution of Cr atoms and vacancies is explored over a range of temperatures and doses using this Monte Carlo model. Sinks and sources for vacancies such as dislocations are accounted for using recombination rates, which are fixed by a specified dislocation density. While variations in input parameters including the temperature and local energy barriers influence the direction and rate of Cr enrichment, all simulation results display a tendency towards saturation in Cr content at the grain boundaries after a sufficiently high dose.
11:30 AM - Q6.4
Phase State and Physical Properties of the Mo-Ru-Ph-Pd Alloys.
Tohru Sugahara 1 , Ken Kurosaki 1 , Aikebaier Yusufu 1 , Hiroaki Muta 1 , Yuji Ohishi 1 , Shinsuke Yamanaka 1 2 , Satoshi Komamine 3 , Toshiki Fukui 3 Show Abstract
1 Graduate School of Engineering, Osaka University, Suita Japan, 2 Research Institute of Nuclear Engineering, Fukui University, Fukui Japan, 3 , Japan Nuclear Fuel Limited, Rokkasho Japan
he Mo-Tc-Ru-Rh-Pd alloys observed in the irradiated nuclear fuels do not dissolve in nitric acid under the reprocessing process and exist as the undissolved residues. The chemical and physical properties of the alloys should be understood to evaluate the safty and economy of the reprocessing process. However, the properties of the alloys have scarcely been reported. In the present study, we investigated the phase state and the physical properties of the alloys composed of Mo, Ru, Rh, Pd, and Mn or Re, in which Mn and Re are representative of Tc.Three kinds of alloys were prepared: Mo2.5M0.5Ru5RhPd (M = Mo, Mn, Re), where M represents Tc. The alloys were prepared by an arc melting technique. The crystal and micro structures were characterized by means of XRD, SEM, and EDX. The lattice parameters were determined by Rietveld refinement, assuming P63/mmc (No. 194) space groups. The temperature dependences of the electrical resistivity and the thermal conductivity were examined. The elastic moduli and the hardness at room temperature were also examined. The high-temperature stability of the alloys under vaious conditions were investigated by TG-DTA analysis.The XRD data indicated that all the samples were almost single phase. The XRD patterns of the alloys fit to the hexiagonal ε phase. The calculated lattice parameters of Mo3Ru5RhPd, Mo2.5Mn0.5Ru5RhPd, and Mo2.5Re0.5Ru5RhPd were a = 2.71 Å, c = 4.28 Å, a = 2.77 Å, c = 4.42 Å, and a = 2.75 Å, c = 4.39 Å, respectively. The electrical, thermal, and mechanical properties as well as the phase state at high temperature of the alloys will be discussed.
11:30 AM - Q6.5
Modelling of WTa and WV alloys for Fusion Applications: Phase Stability, Short-range Order and Irradations Defect Properties.
Marek Muzyk 1 2 , Duc Nguyen-Manh 2 , Krzysztof Kurzydlowski 1 , Nadine Baluc 3 , Sergei Dudarev 2 Show Abstract
1 Faculty of Materials Science and Engineering, Warsaw University of Technology, Warsaw Poland, 2 Theory and Modelling Department, EURATOM/CCFE Fusion Association, Abingdon United Kingdom, 3 Centre of Research in Plasma Physics, Association EURATOM - Swiss Confederation, Federal Institute of Technology, Lausanne Switzerland
Predictions of material properties from atomic-scale simulation is a promising tool for design of materials tailored for specific applications. These simulations provide information about the phase stability of materials as a function of chemical composition, as well as about the formation energies of radiation defects.In our work, we have performed calculations of phase stability, formation energies of radiation defects and the Rice-Thomson criterion of tungsten-based alloys W-Ta and W-V. This alloys are candidate for DEMO divertor applications because of their high melting point and expected improved ductility and fracture toughness in comparison with tungsten. Using ab-initio calculations, we compared enthalpies of mixing for large sets (~100 structures) of alloy configurations, considering several alternative ordered structures corresponding to the same chemical composition. In this way we have identified the lowest energy intermetallic compounds, which should form at low temperatures, and calculated the effective inter-atomic interactions. Using Monte-Carlo calculations, we calculated the temperature of order-disorder phase transformations for these alloys. However, the predicted temperature of order-disorder phase transformations is relatively low and at high temperature it is found that the short-range order is present for both alloys. Ab-initio calculations also show that vanadium atoms strongly trap self-interstitial atom defects in W-V alloys, whereas Ta atoms in W-Ta alloys have very little effect on either the formation energy or thermally activated mobility of self-interstitial atom defects. Using the Rice-Thomson criterion as the screening parameter for identifying ductilizing additives, it is predicted that alloys of more than 25% of V or Ta would lead to brittle to ductile transition from pure tungsten. This work is supported by European Fusion Development Agreement (EFDA) within a collaborative project between the three associations (Polish, CCFE and Swiss).
11:30 AM - Q6.6
The Brittle to Ductile Transition Temperature in W-Ta Alloys for Structural Applications in Nuclear Fusion.
David Armstrong 1 , Steve Roberts 1 , Michael Rieth 2 Show Abstract
1 Department of Materials, University of Oxford, Oxford United Kingdom, 2 Institute for Materials Research I, KIT: Karlsruhe Institute of Technology, Karlsruhe Germany
There is a great deal of interest in using tungsten alloys as both critical plasma facing components and for structural high heat load applications in future nuclear fusion power plants in the divertor. For this to occur, the mechanical properties, and how they change under the extreme conditions faced in the power plant must be understood. Of particular importance is the brittle to ductile transition (BDT). Recently much work has been performed in Oxford using small scale 4 point mechanical tests to measure the brittle to ductile transition temperature (BDTT) in a range of pure tungsten materials. Tests have found that the BDTT temperature of tungsten can range from x to y, depending on the purity and manufacturing process. In an effort to lower the BDTT, tungsten – tantalum alloys have been produced by powder metallurgy followed by forging; this results in a highly textured sample, with a fine elongated grain structure. Four point bend tests have been performed on specimens of W5wt%Ta alloy, at temperatures from 290K to 1470K, at a strain rate of 3.3x10-5s-1. These specimens were spark machined from a larger sample and polished to a size of 900μm x 900μm x12mm. The specimens were cut from a plane which is normal to the forging direction with the test direction parallel to the forging direction.Samples tested at temperatures up to 670K were seen to fracture in a brittle manner at a stress of approximately 1GPa; however the fracture plane tends to follow a path along the long axis of the bar, rather than straight through as seen in pure W alloys. EBSD shows that the path follows specific grain boundaries, which are weaker than the grain interiors and the majority of the grain boundaries. At temperatures above 670K the fracture mode is seen to change from purely brittle to a mixed mode, with a yield stress of 1GPa and failure stress of 1.5GPa. Failure occurs by delamination of the specimen along the same types of boundaries as those seen to fail in a brittle manner at lower temperatures, and significant plastic deformation is seen to occur before failure. SEM shows a fracture surface characterised by regions of ductile deformation interspaced with regions which have failed by intergranular fracture. These results are in good agreement with work carried out on the same materials using high temperature Charpy testing which has shown brittle intergranular fracture at temperatures 273K - 1100K. The cause of the embrittlement is now being further investigated with TEM-EDX. First results indicate that the segregation of oxygen to grain boundaries is the major cause.
11:30 AM - Q6.7
Evaluation of Adhesive Strength Between Vanadium Alloys and Yttrium Oxide by Means of a Laser Shock Method.
Manabu Satou 1 Show Abstract
1 Mechanical Engineering, Hachinohe Institute of Technology, Hachinohe, Aomori, Japan
Abstract Body: Adhesive strength between V-4Cr-4Ti type alloys and yttrium oxide layer made by plasma spray was evaluated by means of a laser shock method, which used pulse laser to generate shock wave that created tensile stress inside the specimen. Typical strength of the layer was evaluated to be about 400 MPa so far. Detailed observation of the exfoliation behavior was carried out to identify the weakest interface of the coating layer. Several modes of the exfoliation behavior were categorized after cross sectional observation. It was found that the adhesive strength of the layer evaluated by the laser shock method had some uncertainty due to the thickness of the coating layer.
11:30 AM - Q6.8
Irradiation Induced Hardening and Void Swelling in Extra High Purity Ni-base Superalloys under Multi-ion Irradiation.
Gwangho Kim 1 , Kiyoyuki Shiba 1 , Tomotsugu Sawai 1 , Ikuo Ioka 1 , Kiyoshi Kiuchi 1 , Jumpei Nakayama 2 Show Abstract
1 , Japan Atomic Energy Agency, Ibaraki Japan, 2 , Kobe Steel, Ltd., Hyogo Japan
In the present work, irradiation performances of extra high purity (EHP) Ni-base superalloys designed as the MA doped MOX fuel claddings for the sodium cooled fast breeder reactor were studied in terms of microstructural changes, irradiation induced hardening and void swelling with respect to multi-ion irradiation by using Takasaki Ion Accelerators for Advanced Radiation Application (TIARA) facility. In this alloy, impurities, such as C, O, N, P, S were reduced less than 100 ppm in total to improve workability, irradiation embrittlement, inter-granular corrosion and weldability. Two types of alloys (Fe-43Ni-20Cr-1.5Al-1.5Ti and Fe-43Ni-25Cr-10W-2.7Si) that are strengthened by precipitates of the ordered γ’ phase or the G phase were used for the ion irradiation experiments. Single (Ni3+) and triple (Ni3+, He+ and H+) beam irradiation were conducted up to ~90 dpa, ~90 appmHe, and ~1350 appmH at 675K and 825K to evaluate the effect of radiation damage and the effect of transmuted gaseous elements. Irradiation induced hardening and void swelling were evaluated and compared with type 316 stainless steel.Other microstructural features, such as precipitation stability, dislocation structures were also evaluated to discuss irradiation resistance.
Tuesday PM, November 30, 2010
Room 208 (Hynes)
2:30 PM - Q7.1
In Situ Characterization of Lattice Structure Evolution during Phase Transformation of Zr-2.5Nb.
Kun Yan 1 2 , David Carr 1 , Mark Reid 2 , Andrew Studer 1 , Robert Harrison 1 , Rian Dippenaar 2 , Klaus-Dieter Liss 1 Show Abstract
1 , Australian Nuclear Science and Technology Organisation, Menai, New South Wales, Australia, 2 , University of Wollongong, Wollongong, New South Wales, Australia
The α-β phase transformation behavior of Zr-2.5Nb has been characterized in real time during an in-situ neutron diffraction experiment. The Zr-2.5Nb material in the current study, at room temperature, consists of α-Zr phase (hcp) and two β phases (bcc), a Nb rich β-Nb phase and retained, Zr rich, β-Zr(Nb) phase. Other than the Burgers and Potter orientation relationships between (002) hcp and (110) bcc, this is the first time the transformation from retained Zr rich β-Zr(Nb) phase to β−Zr with increasing temperature has been reported in the literature. It is suggested that this is related to a dynamic equilibrium of the solubility of Nb atoms in the Zr bcc unit cells.
2:45 PM - Q7.2
Assessment of the Effect of Irradiation Temperature on the Mechanical Anisotropy of the Zr Ion Irradiated Zr-2.5%Nb.
Bipasha Bose 1 , Robert Klassen 1 Show Abstract
1 Mechanical and Materials Engineering, The University of Western Ontario, London, Ontario, Canada
Pressure tubes employed in CANDU nuclear reactors are made of extruded and cold drawn Zr-2.5%Nb. Here we present new information on the effect of Zr ion irradiation, as a simulation of neutron irradiation, on the anisotropy of the local plastic deformation of this pressure tube material. Polished samples, aligned normal to the transverse (TN), axial (AN) and radial (RN) directions of the pressure tube, were irradiated at 25 C and at 300 C to assess the effect of thermal recovery during irradiation on the accumulation of irradiation damage. The samples were irradiated with 8.5 MeV Zr ions. The accumulated irradiation damage ranged with depth within the sample and reached a maximum of 30 displacements per atom (dpa) at a depth of about 2.0 μm. Constant-load pyramidal indentation creep tests were performed at 25 C, at indentation depths from 0.1 to 2.0 μm, before and after ion irradiation. The average indentation stress was found to increase with decreasing indentation depth and with increasing levels of ion irradiation. The ratio of the average indentation stress on TN plane relative to that on the AN and RN planes was found to be 1.3 and 1.2 respectively for the non-irradiated material. These ratios were reduced to 1.04 and 1.08 respectively after Zr ion irradiation at 25 C and reduced to 1.12 and 1.14 after irradiation at 300 C. The change in indentation stress as a result of irradiation damage decreases with increasing resolved basal pole fraction in the indentation direction. This suggests that Zr ion irradiation has a greater effect on blocking the movement of dislocations on prismatic slip systems compared to pyramidal slip systems in the Zr-2.5%Nb pressure tubing.The average activation energy of the indentation creep rate and hence the, activation energy of the obstacles that limit the rate of dislocation glide, was independent of indentation direction but increased with increasing levels of irradiation damage. This indicates that the ion irradiation introduces considerable point defects which act as obstacle to dislocation glide. This is supported by TEM images obtained from foils extracted from beneath the crept indentations showing that ion irradiation produces small, nanometer size, dislocation loops which act as obstacles to dislocation glide and thus influence both the indentation stress and the activation energy.
3:00 PM - Q7.3
In-situ Diffraction Studies on Zirconium Alloys upon Physical Thermo-mechanic Simulation.
Klaus-Dieter Liss 1 , Kun Yan 1 2 , Saurabh Kabra 1 , David Carr 3 , Robert Harrison 3 , Rian Dippenaar 2 Show Abstract
1 The Bragg Institute, Australian Nuclear Science and Technology Organisation, Lucas Heights, New South Wales, Australia, 2 Faculty of Engineering, University of Wollongong, Wollongong, New South Wales, Australia, 3 Institute of Materials Engineering, Australian Nuclear Science and Technology Organisation, Lucas Heights, New South Wales, Australia
Both neutron and high energy X-rays are penetrating probes for the bulk of materials, however, they have complementary properties which can be exploited. Neutrons sample over a large volume so that good grain statistics is obtained while the synchrotron beam allows to determine information from individual grains. Here we report on Zircaloy-4 and Zr-2.5Nb, which are, due to their neutron transparency and performance under radiation, important structural materials for the nuclear power industries.Specimens have been heated to investigate in-situ by neutron diffraction the phase transformation behavior. Further, tensile tests have been undertaken at high temperatures, probed in-situ and in real time by a synchrotron beam. Details, such as the evolution of the lattice parameter, phase fractions, grain orientation relationships, grain refinement, dynamic recovery and dynamic recrystallization are revealed from the multi-dimensional data.
3:15 PM - Q7.4
Microstructures in Unirradiated U-Pu-Zr metal Fuels for Nuclear Reactors.
Dawn Janney 1 , Rory Kennedy 1 Show Abstract
1 , Idaho National Laboratory, Idaho Falls, Idaho, United States
U-Pu-Zr alloys have been considered as fuel for nuclear reactors since the early days of atomic energy, and are currently being investigated for transmutation of minor actinides (Np, Am, Cm) in a closed fuel cycle. Despite the potential importance of these materials, microstructures in unirradiated U-Pu-Zr alloys have not previously been described in detail. This presentation reports results of scanning electron microscope (SEM) characterization of as-cast microstructures in U-Pu-Zr alloys with compositions including 50U-30Pu-20Zr, 25U-55Pu-20Zr, and 35U-20Pu-45Zr (where numbers indicate wt% of the specified elements).The samples were arc-cast into fused silica tubes with a nominal diameter of ~ 4 mm in an argon-atmosphere glovebox. The samples were then sliced with a slow-speed saw in an air-atmosphere glovebox, embedded in acrylic or epoxy, ground and polished to produce a smooth surface, and coated with a thin layer of sputtered Pd. SEM examination was done with a JEOL 7000F field-emission SEM equipped with energy- and wavelength-dispersive spectrometry. All of the samples contain high-Zr inclusions in a high-actinide matrix. Morphologies and spatial distributions of the inclusions vary; however, many appear similar to the high-Zr inclusions previously reported in published descriptions, for example, binary U-Zr alloys. Rows of sub-parallel, elongated Zr inclusions forming irregular polygons also occur in 35U-20Pu-45Zr. To a first approximation, the matrix in 35U-20Pu-45 Zr appears homogeneous. Samples from two castings of 50U-30Pu-20Zr have significantly different matrix microstructures, with one sample having pronounced variations in U/Zr ratios and the other having areas with what appear to be sub-micron high-actinide particles. The matrix in 25U-55Pu-20Zr shows complex intergrowths of two distinct compositions containing different ratios of U, Pu, and Zr.Previous research on the U-Pu-Zr phase diagram indicates solid solution throughout the entire composition range at the solidus. X-ray diffraction data from the samples to be discussed in this presentation indicate that these samples consist almost entirely of low-temperature phases, suggesting that the present microstructures may have formed by solid-state reactions from an earlier solid solution phase.
Q8: Modeling - He Effects & Void Formation
Tuesday PM, November 30, 2010
Room 208 (Hynes)
4:00 PM - **Q8.1
A Review of Rate Theory Models of Helium Bubble Evolution Under Irradiation.
Nasr Ghoniem 1 Show Abstract
1 Mech & Aerospace, UCLA, Los Angeles, California, United States
Rate theory models of the nucleation and growth of helium-filled cavities in irradiated materials will be reviewed. Classical nucleation theory is inadequate under fusion conditions(high helium-to-dpa ratios) and the usual “mean field”approximation of microstructural growth cannot account for cascade effects. A comprehensive theory of helium-filled cavity evolution under neutron irradiation conditions is formulated based on non-equilibrium statistical mechanics. First, helium diffusion in irradiated materials will be shown to be inter-linked to the nucleation and growth of helium-filled cavities, and that the transport of helium to microstructural features, such as dislocations, cavities, precipitates and grain boundaries is dictated by its interaction with continuously nucleated cavities. The importance of “secondary nucleation”as a result of the dynamic re-solution of helium gas in cavities will be emphasized, as it results in significantly higher concentrations ofhelium bubbles under collision cascade conditions. The transport and nucleation processes of helium are modeled by discrete rate equations (the so-called cluster dynamics), while large sizes are treated with approximate continuum methods. A review of these methods will be given, which include, Fokker-Planck, distribution moments, and various grouping techniques. Under fusion irradiation conditions, where the ratio of helium-to-displacement damage production is high, we show that simplifications arise in solving rate equations, enabling direct comparison with experiments, and the coupling betweenhelium-vacancy clustering inside grains with cavity evolution on grain boundaries. The problem of spatial self-organization of cavities under irradiation is described in terms of a Ginzburg-Landau-type equation and the results are compared to experiments.
4:30 PM - Q8.2
Helium-vacancy Cluster in Single BCC Iron Crystal Lattice.
Ning Gao 1 , Maximo Victoria 1 , Jiachao Chen 1 , Helena Van Swygenhoven 2 Show Abstract
1 NES, Paul Scherrer Institut, Villigen Switzerland, 2 NUM/ASQ, Paul Scherrer Institut, Villigen Switzerland
The properties of clusters formed of one vacancy and with an increasing number of He atoms (nHe-V) is studied in bcc iron with molecular statics (MS) and molecular dynamics (MD) simulations. The binding energies of He to a nHe-V cluster are calculated and compared with DFT predictions. Both DFT and the empirical potential give similar results. The calculated binding energies of the helium atom to the nHe-V cluster decrease first, then slowly increases up to n = 4, it is then almost constant between n = 5 to 15. At n = 16 the binding energy increases strongly again. This last increase is related to the athermal SIA emission in the form of <110> dumbbell which relaxes the restrained nHe-V cluster. A series of near-symmetrical or symmetrical He atom arrangements based on the fcc configuration are created with increasing He content. The work done by the nHe-V cluster as it displaces the surrounding Fe lattice atoms away from initial perfect sites increases continuously with n and exceeds the formation energy of a SIA at n ≥ 6. The local pressure calculation shows that as the number of He atoms into the vacancy increases, the local peak normal stress and shear stress also increased up to about 9 GPa and 4 GPa, respectively.
4:45 PM - Q8.3
Atomistic Study of Helium Bubble Strengthening in Fe.
Yury Osetskiy 1 , David Stewart 1 , Roger Stoller 1 Show Abstract
1 Materials Science and Technology, ORNL, Oak Ridge , Tennessee, United States
In the fusion irradiation environment, helium created by transmutation will play an important role in the response of structural materials to neutron radiation damage. Recently we carried out atomistic simulations to investigate the equilibrium state of He-filled bubbles in Fe and have found that the equilibrium He content is rather low and at a room temperature it is ~1.0 to 0.3 He per vacancy for bubbles of diameters from 1 to 6 nm. We present the results of a study of bubbles as obstacles for edge dislocation motion. We have simulated bubbles with different He content at different temperatures from 100 to 600K. It was found that the resistance bubbles create to dislocation motion depends on their size and He-content. Depending on He content, a bubble can be a weaker or stronger obstacle in comparison with a void of the same size. Equilibrium bubbles were found to be the strongest obstacles. The interaction mechanism also depends on the He content and changes from absorption of atoms into the bubble (underpressurized bubble) to absorption of vacancies (near equilibrium) to emission of interstitials cluster (overpressurized bubble). The results on bubbles are compared with other spherical obstacles such as voids and precipitates.
5:00 PM - Q8.4
Atomistic Study of Helium Bubbles in Fe: Equilibrium State.
David Stewart 1 2 , Yuri Osetsky 1 , Roger Stoller 1 Show Abstract
1 Materials Science and Technology, Oak Ridge National Laboratory, Oak Ridge, Tennessee, United States, 2 Center for Materials Processing, The University of Tennessee, Knoxville, Tennessee, United States
In the fusion irradiation environment, helium created by transmutation will play an important role in the response of structural materials to neutron radiation damage. Recently we have developed a new 3-body potential to describe Fe-He interaction in Fe matrix. We have used here this potential to investigate the equilibrium state of He bubbles embedded into the bcc Fe matrix. We have investigated bubble size, He content and temperature effects. It was found that the equilibrium He content is rather low and at a room temperature it is ~1.0 to 0.3 He per vacancy for bubbles of diameters from 1 to 6 nm. The equilibrium He/vacancy ratio decreases with temperature increase and for bubbles of 6 nm in diameter drops down to <0.2 at 900K. The distribution of He atoms inside bubbles and effects of bubble facet crystallography were also investigated. The results are compared with the capillarity model often used for estimating the equilibrium pressure of He bubbles.
5:15 PM - Q8.5
Mesoscale Simulation of Irradiation-induced Gas Bubbles: Evolution and Impact on Macroscale Properties.
Paul Millett 1 , Anter El-Azab 2 , Michael Tonks 1 Show Abstract
1 Fuels Modeling and Simulation, Idaho National Laboratory, Idaho Falls, Idaho, United States, 2 Department of Scientific Computing, Florida State University, Tallahassee, Florida, United States
Nuclear materials used in fission and fusion applications can be subjected to simultaneously high displacement rates as well as gas production rates. The characteristics of the resultant gas bubble structure, i.e. internal gas density, bubble number density, and bubble distribution, are strongly dependent on the relative defect production rates as well as microstructure (e.g. grain size). Here, we implement a phase-field model capable of capturing multi-component defect (vacancies, self-interstitials, and gas atoms) diffusion, bubble nucleation and growth, and bubble/grain boundary interactions to investigate these processes throughout time and for varying irradiation conditions. Furthermore, we have utilized this simulation capability to develop models of bubble percolation and the effective thermal transport across heterogeneous microstructures. Interestingly, thermal conductivity is found to be strongly dependent on the relative distribution of intergranular versus intragranular bubbles. This research was supported by the Nuclear Energy Advanced Modeling and Simulation (NEAMS) program within DOE-NE.
5:30 PM - Q8.6
Phase Field Modeling of Void Microstructure Evolution and Swelling in Irradiated Metals.
Srujan Rokkam 1 , Anter El-Azab 2 , Jie Deng 2 Show Abstract
1 Mechanical Engineering, Florida State University, Tallahassee, Florida, United States, 2 Department of Scientific Computing, Florida State University, Tallahassee, Florida, United States
Void formation and associated swelling is one of the most technologically relevant and intriguing problems in the design of structural materials for nuclear reactor elements. The conventional approach for modeling the void swelling problem considers the void nucleation and growth stages separately, which are treated as uniform processes in space using classical nucleation theory and rate theory, respectively. Here, we present a phase field model for void formation which treats the void nucleation and growth processes simultaneously in a spatially resolved fashion. Using the principles of irreversible thermodynamics, the point defect fields are described using Cahn-Hilliard type equations. The dynamics of void growth is obtained in terms of the evolution of a non-conserved order parameter, prescribed by phenomenological Allen-Cahn type equation. Point defects generated by atomic displacement cascades are introduced in a localized and segregated fashion using a core-shell type model. The model accounts for mutual recombination of point defects, interactions with extended defects, sinks, effects of applied stress, cascade-induced and thermally induced fluctuations. Using two dimensional solutions, we demonstrate our model capabilities for void nucleation, growth and swelling resulting from cascade-induced defects. Contrary to the populist belief, we observe that the swelling of the material can occur prior to the nucleation of voids, by absorption of interstitials at the material surface. It is observed that void nucleation in fact restricts the amount of swelling. Further, the dependence of swelling on material properties, temperature and dose-rate is investigated. This work was performed as part of a Computational Materials Science Network (CMSN) project supported by the US Department of Energy, Office of Basic Energy Sciences via contract number DEFG02-07ER46367 at Florida State University.