D. Kaoumi

A thermal spike model of grain growth under irradiation

The experimental study of grain growth in nanocrystalline metallic foils under ion irradiation showed the existence of a low-temperature regime (below about 0.15–0.22Tm), where grain growth is independent of the irradiation temperature, and a thermally assisted regime where grain growth is enhanced with increasing irradiation temperature. A model is proposed to describe grain growth under irradiation in the temperature-independent regime, based on the direct impact of the thermal spikes on grain boundaries. In the model, grain-boundary migration occurs by atomic jumps, within the thermal spikes, biased by the local grain-boundary curvature driving. The jumps in the spike are calculated based on Vineyard’s analysis of thermal spikes and activated processes using a spherical geometry for the spike. The model incorporates cascade structure features such as subcascade formation, and the probability of subcascades occurring at grain boundaries. This results in a power law expression relating the average grain ...

In situ transmission electron microscopy and ion irradiation of ferritic materials

The intermediate voltage electron microscope‐tandem user facility in the Electron Microscopy Center at Argonne National Laboratory is described. The primary purpose of this facility is electron microscopy with in situ ion irradiation at controlled sample temperatures. To illustrate its capabilities and advantages a few results of two outside user projects are presented. The motion of dislocation loops formed during ion irradiation is illustrated in video data that reveals a striking reduction of motion in Fe‐8%Cr over that in pure Fe. The development of extended defect structure is then shown to depend on this motion and the influence of nearby surfaces in the transmission electron microscopy thin samples. In a second project, the damage microstructure is followed to high dose (200 dpa) in an oxide dispersion strengthened ferritic alloy at 500°C, and found to be qualitatively similar to that observed in the same alloy neutron irradiated at 420°C. Microsc. Res. Tech., 2009.

TEM Characterization of Crept and Irradiated Nano-structured Ferritic Alloys

The past ten years or so have seen the development of an exciting new class of mechanically alloyed (MA) nano-structured ferritic alloys (NFA) with outstanding mechanical properties that are mostly due to the presence of high concentrations (>10 23 m -3 ) of Ti-, Y-, and O-enriched nano-clusters (NC). Because NC may promote point defect recombination and trap transmutation-produced He in small clusters, NFA have the potential to be highly resistant to radiation damage in fission and fusion environments [1,2], and thus are being characterized following neutron and ion irradiation. Energy-filtered transmission electron microscopy (EFTEM) performed at 300 kV on a LaB6 Philips CM30 equipped with a Gatan imaging filter (GIF) has been especially beneficial for imaging NC. In particular, Fe-M jump-ratio images produced from component images recorded with 10-eV slits at energy losses of 46 and 62 eV reliably reveal NC in dark contrast. Such images are insensitive to surface oxide films or modest surface contamination and for sufficiently thin regions (<50 nm) 2-nm diameter NC are visible [3]. Additional EFTEM elemental mapping (e.g. O, Ti-L23, Cr-L23) has also been usefully applied to NFA, and focused-ion-beam (FIB) lift-out specimens have been used to good advantage [2]. Fabrication of an Fe-14.2wt.%Cr-1.95%W-0.22%Ti-0.25%Y2O3 NFA, designated 14YWT, has been described elsewhere, as have the contributions of TEM to help optimize material processing parameters [3,4]. It was also previously shown that NC in 14YWT are not detectably changed by tensile testing at 25 and 700°C with total strains of up to 39% [5] and that in MA957 (an INCO-patented Fe-14wt%Cr-1%Ti-0.3%Mo-0.27%Y2O3 NFA) neutron irradiated at 500°C to 9 displacements per atom (dpa) and with ~380 appm He, the diameter (~3 nm) and concentration (~4 x 10 23 m -3 ) of the NC differ little from those of unirradiated MA957 [1,2].

High Fidelity Ion Beam Simulation of High Dose Neutron Irradiation

The objective of this proposal is to demonstrate the capability to predict the evolution of microstructure and properties of structural materials in-reactor and at high doses, using ion irradiation as a surrogate for reactor

Effect of dose on irradiation-induced loop density and Burgers vector in ion-irradiated ferritic/martensitic steel HT9

ABSTRACT Samples of F/M steel HT9 were irradiated to 20 dpa at 420°C, 440°C and 470°C in a transmission electron microscope with 1 MeV Kr ions so that the microstructure evolution could be followed in situ and characterised as a function of dose. Dynamic observations of irradiation-induced defect formation and evolution were made at the different temperatures. Irradiation-induced loops were characterised in terms of their Burgers vector, size and density as a function of dose and similar observations and trends were found at the three temperatures: (i) both a/2 <111> and a <100> loops are observed; (ii) in the early stage of irradiation, the density of irradiation-induced loops increases with dose (0–4 dpa) and then decreases at higher doses (above 4 dpa), (iii) the dislocation line density shows an inverse trend to the loop density with increasing dose: in the early stages of irradiation, the pre-existing dislocation lines are lost by climb to the surfaces while at higher doses (above 4 dpa), the build-up of new dislocation networks is observed along with the loss of the radiation-induced dislocation loops to dislocation networks; (iv) at higher doses, the decrease of number of loops affects more the a/2 <111> loop population; the possible loss mechanisms of the a/2 <111> loops are discussed. Also, the ratio of a <100> to a/2 <111> loops is found to be similar to cases of bulk irradiation of the same alloy using 5 MeV Fe2+ ions to similar doses of 20 dpa at similar temperatures.

Microstructural and Nanomechanical Characterization of In-Situ He Implanted and Irradiated fcc Materials

Nuclear Materials research focusing on the degradation of materials utilized in nuclear environments has been studied for decades. This research aims to assess the microstructural changes and therefore mechanical property degradation as a function of radiation dose at representative temperatures and environments. Traditionally specimens are exposed to reactor or ion beam irradiation and subsequently examined using transmission electron microscopy, atom probe tomography, X-ray techniques and mechanical testing. However, all these techniques are mostly conducted ex-situ and one only observes the resulting changes but not on the same sample or same location making interpretations sometimes challenging. Following the defect development has been achieved utilizing tools like the Intermediate Voltage Electron Microscopy (IVEM) facility or other similar in-situ irradiation facilities [1]. Modern He ion beam microscopes also allow direct implantation of He in samples on a very localized level allowing localized He implantation on specific regions of interest on the same TEM foil [2]. While these are a very useful tools the damage evolution has only been followed on TEM foils and no mechanical data has been extracted from the irradiated samples. Ideally one would be able to extract a change in mechanical property associated with the microstructural changes. Recent advantages in nanomechanical testing and in-situ ion beam irradiation have the potential to obtain both microstructural data as well as mechanical property data on the same sample. This work attempts to perform the entire study in-situ spanning from He implantation to displacement damage and subsequent mechanical property evaluation to follow the defect development from the beginning to its implications for mechanical properties. It is anticipated that this procedure developed here will allow to establish a process producing data for model benchmarking bridging the gap between modeling and experiments on the same length-scale.

Microstructure Characterization of Ion-irradiated Ferritic/Martensitic HT9 Steel

HT9 is a 12Cr Ferritic/Martensitic (F/M) steel considered as a promising candidate for structural and cladding applications in Generation IV reactors [1]. The chemical composition of the alloy is given in Table 1. The harsh service conditions in Gen IV reactors require that the microstructural response to irradiation of the candidate structural alloys be investigated and understood to qualify them. For that matter, a series of ion irradiations were done. Bulk HT9 specimens were irradiated using 5 MeV Fe ions to 20 displacements per atom (dpa) at 600 nm depth with a dose rate approximately to 5×10 dpa/s, at irradiation temperatures of 420, 440 and 470°C (with a variation of ± 5°C). The temperature was monitored using an infrared camera and four attached Type J thermocouples. For post-irradiation characterization, TEM specimens were firstly prepared by the FIB lift-out method using a FEI Quanta focused ion beam (FIB) instrument. ChemiSTEM characterizations were then conducted on FIB laminas using a FEI Titan 80-300 probe aberration corrected microscope. ChemiSTEM characterization was also conducted on as-received HT9 prior to ion irradiation. Only pre-existed M23C6 type carbides and V-rich nitride precipitates were observed in the as-received condition. In contrast, Ni-Si-Mn rich precipitates (also known as G phase) were found in HT9 irradiated to 20 dpa at 420, 440 and 470°C, as shown in Figure 1. Radiation-induced Ni segregation was also observed at grain boundaries, which is highlighted by white arrows in Figure 2. In addition, the G phase precipitates were found to nucleate heterogeneously along lath grain boundaries, as indexed by red arrows in Figure 2. The observed results indicated that, under self-ion irradiation, alloying elements such as Ni, Si and Mn segregate at defect sinks, which become thus favourable nucleation sites and promote the radiation-induced G phase precipitation. While radiation induced precipitation and segregation in neutron irradiated F/M HT9 have been widely reported in the literature, similar investigations under ion irradiation have been more scarce [2,3,4]. In fact, this study serves to generate baseline data on ion irradiation effects on F/M HT9 in an effort to learn how to more accurately choose ion-irradiation experimental conditions to emulate the irradiated microstructures and effects observed under neutron irradiation. Ion irradiations were also carried out in the same alloy at similar temperatures in-situ in a TEM using 1MeV Kr++ ions so that the microstructure characterized in-situ in the TEM can be compared with the microstructure achieved on the same alloys using self-ion irradiation on bulk samples. The focus of the comparison is put on the size and density of dislocation loops induced by irradiation, as well as dislocation loop burgers vector determination. The in-situ experiments provide data on the kinetics of irradiation induced defect formation and evolution, and on the damage spatial correlation with the preexisting microstructure, and thus can help understand how the microstructures observed ex-situ in the bulk samples have developed for, in these latter cases, only snapshots are available at the limited doses. By comparing the ex-situ and in-situ irradiation it is also possible to substantiate the free surface effect on the radiation induced microstructure. The presentation will also report such comparison.

Using In-Situ TEM to Characterize the Microstructure Evolution of Metallic Systems under External Solicitation

Studying materials under external stimulus such as irradiation and/or mechanical stress can be difficult because of the lack of kinetics information, since usually samples are examined ex situ (e.g. after irradiation or after mechanical testing) so that only discrete snapshots of the process are available. Given the dynamic nature of the phenomena, direct in situ observation is often necessary to better understand the mechanisms, kinetics and driving forces of the processes involved. For this matter, using in situ Transmission Electron Microscopy (TEM) can be of great help[1]. Indeed, the spatial resolution of the TEM makes it an invaluable tool in which one can continuously track the real-time response of the microstructure to external stimuli, which can help discover and quantify the fundamental rate-limiting microscopic processes and mechanisms governing the macroscopic properties. In this presentation, two examples will be given which show how the technique can be used for nuclear engineering applications. (i) In-situ straining experiments in the TEM is applied to investigate deformation mechanisms in Ni-based alloys (Inconel 617 and Haynes 230) which are candidate materials for the heat exchanger in the GEN-IV Very High Temperature nuclear Reactor. In addition to showing dislocation dynamics under tensile strain, it also allows to follow crack propagation as it proceeds in the material. (ii) In-situ Ion-irradiation in the TEM has proven a very good tool for studying the basic mechanisms of radiation damage formation and evolution as a function of dose, dose rate, temperature and ion type.

Microstructure and Property Evolution in Advanced Cladding and Duct Materials Under Long-Term and Elevated Temperature Irradiation: Modeling and Experimental Investigation

The in-service degradation of reactor core materials is related to underlying changes in the irradiated microstructure. During reactor operation, structural components and cladding experience displacement of atoms by collisions with neutrons at temperatures at which the radiation-induced defects are mobile, leading to microstructure evolution under irradiation that can degrade material properties. At the doses and temperatures relevant to fast reactor operation, the microstructure evolves by dislocation loop formation and growth, microchemistry changes due to radiation-induced segregation, radiation-induced precipitation, destabilization of the existing precipitate structure, and in some cases, void formation and growth. These processes do not occur independently; rather, their evolution is highly interlinked. Radiationinduced segregation of Cr and existing chromium carbide coverage in irradiated alloy T91 track each other closely. The radiation-induced precipitation of Ni-Si precipitates and RIS of Ni and Si in alloys T91 and HCM12A are likely related. Neither the evolution of these processes nor their coupling is understood under the conditions required for materials performance in fast reactors (temperature range 300-600°C and doses beyond 200 dpa). Further, predictive modeling is not yet possible as models for microstructure evolution must be developed along with experiments to characterize these key processes and provide tools for extrapolation. To extend the range of operation of nuclear fuel cladding and structural materials in advanced nuclear energy and transmutation systems to that required for the fast reactor, the irradiation-induced evolution of the microstructure, microchemistry, and the associated mechanical properties at relevant temperatures and doses must be understood. Predictive modeling relies on an understanding of the physical processes and also on the development of microstructure and microchemical models to describe their evolution under irradiation. This project will focus on modeling microstructural and microchemical evolution of irradiated alloys by performing detailed modeling of such microstructure evolution processes coupled with well-designed in situ experiments that can provide validation and benchmarking to the computer codes. The broad scientific and technical objectives of this proposal are to evaluate the microstructure and microchemical evolution in advanced ferritic/martensitic and oxide dispersion strengthened (ODS) alloys for cladding and duct reactor materials under long-term and elevated temperature irradiation, leading to improved ability to model structural materials performance and lifetime. Specifically, we propose four research thrusts, namely Thrust 1: Identify the formation mechanism and evolution for dislocation loops with Burgers vector of a and determine whether the defect microstructure (predominately dislocation loop/dislocation density) saturates at high dose. Thrust 2: Identify whether a threshold irradiation temperature or dose exists for the nucleation of growing voids that mark the beginning of irradiation-induced swelling, and begin to probe the limits of thermal stability of the tempered Martensitic structure under irradiation. Thrust 3: Evaluate the stability of nanometer sized Y- Ti-O based oxide dispersion strengthened (ODS) particles at high fluence/temperature. Thrust 4: Evaluate the extent to which precipitates form and/or dissolve as a function of irradiation temperature and dose, and how these changes are driven by radiation induced segregation and microchemical evolutions and determined by the initial microstructure.