M. A. Kirk

University of Oxford, Oxford, England, United Kingdom

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Publications (184)207.95 Total impact

  • Journal of Nuclear Materials 09/2014; · 2.02 Impact Factor
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    ABSTRACT: Oxide Dispersion Strengthened (ODS) reduced activation ferritic steels are promising candidate materials for structural components of both nuclear fission and fusion reactors. However, when irradiated with energetic particles, they may suffer changes on their microstructures that degrade their mechanical performance. In-situ transmission electron microscopy studies on ion-irradiated ODS steels can give remarkable insights into fundamental aspects of radiation damage allowing dynamic observations of defect formation, mobilities, and interactions during irradiation. In this investigation, a commercially available PM2000 ODS steel was in-situ irradiated with 150 KeV Fe+ at room temperature and 700°C. These experiments showed that the oxide nanoparticles in these steels remain stable up to the higher irradiation dose (~ 1.5 dpa), and that these particles seem to be effective sinks for irradiation induced defects.
    Journal of Physics Conference Series 06/2014; 522(1):012032.
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    ABSTRACT: Phase stability of Ni3(Al, Ti) precipitates in Inconel X-750 under cascade damage was studied using heavy ion irradiation with transmission electron microscope (TEM) in situ observations. From 333 K to 673 K (60 °C to 400 °C), ordered Ni3(Al, Ti) precipitates became completely disordered at low irradiation dose of 0.06 displacement per atom (dpa). At higher dose, a trend of precipitate dissolution occurring under disordered state was observed, which is due to the ballistic mixing effect by irradiation. However, at temperatures greater than 773 K (500 °C), the precipitates stayed ordered up to 5.4 dpa, supporting the view that irradiation-induced disordering/dissolution and thermal recovery reach a balance between 673 K and 773 K (400 °C and 500 °C). Effects of Ti/Al ratio and irradiation dose rate are also discussed.
    Metallurgical and Materials Transactions A 04/2014; 45a(6). · 1.73 Impact Factor
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    ABSTRACT: In the current investigation, TEM in-situ heavy ion (1MeV Kr2þ) irradiation with helium pre-injected at elevated temperature (400 �C) was conducted to simulate in-reactor neutron irradiation induced damage in CANDU spacer material Inconel X-750, in an effort to understand the effects of helium on irradiation induced cavity microstructures. Three different quantities of helium, 400 appm, 1000 appm, and 5000 appm, were pre-injected directly into TEM foils at 400 �C. The samples containing helium were then irradiated in-situ with 1MeV Kr2þ at 400 �C to a final dose of 5.4 dpa (displacement per atom). Cavities were formed from the helium injection solely and the cavity density and size increased with increasing helium dosage. In contrast to previous heavy ion irradiations with cold pre-injected helium, heterogeneous nucleation of cavities was observed. During the ensuing heavy ion irradiation, dynamical observation showed noticeable size increase in cavities which nucleated close to the grain boundaries. A “bubble-void” transformation was observed after Kr2þ irradiation to high dose (5.4 dpa) in samples containing 1000 appm and 5000 appm helium. Cavity distribution was found to be consistent with in-reactor neutron irradiation induced cavity microstructures. This implies that the distribution of helium is greatly dependent on the injection temperature, and helium pre-injection at high temperature is preferred for simulating the migration of the transmutation produced helium.
    Journal of Applied Physics 03/2014; · 2.21 Impact Factor
  • D R Mason, X Yi, M A Kirk, S L Dudarev
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    ABSTRACT: Using _in situ_ transmission electron microscopy (TEM), we have observed nanometre scale dislocation loops formed when an ultra-high-purity tungsten foil is irradiated with a very low fluence of self-ions. Analysis of the TEM images has revealed the largest loops to be predominantly of prismatic 1/2<111> type and of vacancy character. The formation of such dislocation loops is surprising since isolated loops are expected to be highly mobile, and should escape from the foil. In this work we show that the observed size and number density of loops can be explained by the fact that the loops are _not_ isolated - the loops formed in close proximity in the cascades interact with each other and with vacancy clusters, also formed in cascades, through long-range elastic fields, which prevent the escape of loops from the foil. We find that experimental observations are well reproduced by object Kinetic Monte Carlo simulations of evolution of cascades _only_ if elastic interaction between the loops is taken into account. Our analysis highlights the profound effect of elastic interaction between defects on the microstructural evolution of irradiated materials.
    Journal of Physics Condensed Matter 02/2014; 26(37). · 2.22 Impact Factor
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    ABSTRACT: Heavy ion irradiation induced damage in Inconel X-750 at low temperatures (60–400 °C) has been reported in our previous study. In the current investigation, the microstructure evolution and phase change during heavy (1 MeV Kr2+) irradiation at elevated temperatures (500 °C and 600 °C) were characterized under in situ observation of intermediate voltage electron microscope (IVEM) at Argonne National Laboratory. For each temperature, defect analyses using the weak beam dark field method were carried out at several doses, up to 5.4 dpa. Small defects (<5 nm) yielded from high temperature irradiation comprise mainly stacking fault tetrahedras (SFTs), small ⅓ 〈1 1 1〉 and ½ 〈1 1 0〉 type dislocation loops. Large interstitial Frank loops were observed and a clear characteristic for growth of loops was video-captured. Unfaulting of interstitial Frank loops was observed. The number density of the defects saturated at a relatively low dose of 0.68 dpa. No obvious change of defect fraction was found with increasing dose, but more complex dislocation structures formed at higher doses. In contrast to low temperature irradiation, the primary strengthening phase γ′ was found to be stable during irradiation at temperatures >500 °C and was not disordered up to 5.4 dpa. No cavities were observed after the irradiation even at 600 °C.
    Journal of Nuclear Materials. 01/2014; 445(s 1–3):227–234.
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    ABSTRACT: Defect sinks, such as grain boundaries and phase boundaries, have been widely accepted to improve the irradiation resistance of metallic materials. However, free surface, an ideal defect sink, has received little attention in bulk materials as surface-to-volume ratio is typically low. Here by using in situ Kr ion irradiation technique in a transmission electron microscope, we show that nanoporous (NP) Ag has enhanced radiation tolerance. Besides direct evidence of free surface induced frequent removal of various types of defect clusters, we determined, for the first time, the global and instantaneous diffusivity of defect clusters in both coarse-grained (CG) and NP Ag. Opposite to conventional wisdom, both types of diffusivities are lower in NP Ag. Such a surprise is largely related to the reduced interaction energy between isolated defect clusters in NP Ag. Determination of kinetics of defect clusters is essential to understand and model their migration and clustering in irradiated materials. T he successful development of advanced nuclear reactors calls for the discovery of advanced materials that can endure unprecedented neutron irradiation damage to hundreds of displacements-per-atom (dpa) 1–3 . A high density of irradiation-induced defect clusters, including dislocation loops and networks, voids, bubbles and stacking fault tetrahedra (SFTs), can significantly degrade mechanical properties of materials 4–7 . Several types of defect sinks have been explored to achieve enhanced radiation tolerance, such as high-angle grain boundaries (GBs) 8–13 , immiscible interfaces in nanolayer composites 14–17 , twin boundaries 18,19 and phase boundaries 20,21 . Metallic nanoporous (NP) materials with large surface-to-volume ratios have applications for energy storage, catalysts, filters and gas sensors 22 . Their mechanical, catalytic and optical properties have been widely investi-gated 23–25 . Lee et al. 26 reported mechanical strength of NP Au and suggested that NP metals could be used as high strength, low density materials. Kucheyev et al. 27 concluded that the pronounced time-dependent creep of NP silica at room temperature was attributed to the stress corrosion fracture of nanoscale ligaments. The impact of free surface on irradiation-induced damage in bulk materials has also been studied. In general a larger number of defect clusters (mostly vacancy loops) were observed near-surface compared to those in materials interior (inside bulk materials) 28–31
    01/2014; 4:3737.
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    ABSTRACT: Defect sinks, such as grain boundaries and phase boundaries, have been widely accepted to improve the irradiation resistance of metallic materials. However, free surface, an ideal defect sink, has received little attention in bulk materials as surface-to-volume ratio is typically low. Here by using in situ Kr ion irradiation technique in a transmission electron microscope, we show that nanoporous (NP) Ag has enhanced radiation tolerance. Besides direct evidence of free surface induced frequent removal of various types of defect clusters, we determined, for the first time, the global and instantaneous diffusivity of defect clusters in both coarse-grained (CG) and NP Ag. Opposite to conventional wisdom, both types of diffusivities are lower in NP Ag. Such a surprise is largely related to the reduced interaction energy between isolated defect clusters in NP Ag. Determination of kinetics of defect clusters is essential to understand and model their migration and clustering in irradiated materials.
    Scientific Reports 01/2014; 4:3737. · 5.08 Impact Factor
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    ABSTRACT: In order to understand radiation damage in the nickel based superalloy Inconel X-750 in thermal reactors, where (n, α) transmutation reaction also occurred in addition to fast neutron induced atomic displacement, heavy ion (1 MeV Kr2+) irradiation with pre-injected helium was performed under in-situ observations of an intermediate voltage electron microscope at Argonne National Laboratory. By comparing to our previous studies using 1 MeV Kr2+ irradiation solely, the pre-injected helium was found to be essential in cavity nucleation. Cavities started to be visible after Kr2+ irradiation to 2.7 dpa at ≥200 °C in samples containing 200 appm, 1000 appm, and 5000 appm helium, respectively, but not at lower temperatures. The cavity growth was observed during the continuous irradiation. Cavity formation appeared along with a reduced number density of stacking fault tetrahedra, vacancy type defects. With higher pre-injected helium amount, a higher density of smaller cavities was observed. This is considered to be the result of local trapping effect of helium which disperses vacancies. The average cavity size increases with increasing irradiation temperatures; the density reduced; and the distribution of cavities became heterogeneous at elevated temperatures. In contrast to previous characterization of in-reactor neutron irradiated Inconel X-750, no obvious cavity sink to grain boundaries and phase boundaries was found even at high doses and elevated temperatures. MC-type carbides were observed as strong sources for agglomeration of cavities due to their enhanced trapping strength of helium and vacancies.
    Journal of Applied Physics 01/2014; 115(10):103508-103508-8. · 2.21 Impact Factor
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    ABSTRACT: The in situ ion irradiation technique has been employed to elucidate irradiation damage in the dual phase Zr-Excel alloy. 1 MeV Kr ion irradiation experiments were conducted at different temperatures ranging from 100 °C to 400 °C. Damage microstructures have been characterized by transmission electron microscopy in both the alpha (α) and beta (β) phases after a maximum dose of 10 dpa at different temperatures. Several important observations including low temperature 〈c〉-component loop formation, and irradiation induced omega (ω) phase precipitation have been reported. In situ irradiation provided an opportunity to observe the nucleation and growth of basal plane 〈c〉-component loops and irradiation induced dissolution of secondary phase precipitates at the same time. It has been shown that under Kr ion irradiation the 〈c〉-component loops start to nucleate and grow above a threshold dose, as has been observed for neutron irradiation. Furthermore, the role of temperature, material composition and pre-irradiation microstructure has been discussed in detail.
    Journal of Nuclear Materials 10/2013; 441(1-3):138-151. · 2.02 Impact Factor
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    Scripta Materialia 09/2013; 69(5):385–388. · 2.82 Impact Factor
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    ABSTRACT: Monolithic Ag and Ni films and Ag/Ni multilayers with individual layer thickness of 5 and 50 nm were subjected to in situ Kr ion irradiation at room temperature to 1 displacement-per-atom (a fluence of 2 × 1014 ions/cm2). Monolithic Ag has high density of small loops (4 nm in diameter), whereas Ni has fewer but much greater loops (exceeding 20 nm). In comparison, dislocation loops, ∼4 nm in diameter, were the major defects in the irradiated Ag/Ni 50 nm film, while the loops were barely observed in the Ag/Ni 5 nm film. At 0.2 dpa (0.4 × 1014 ions/cm), defect density in both monolithic Ag and Ni saturated at 1.6 and 0.2 × 1023/m3, compared with 0.8 × 1023/m3 in Ag/Ni 50 nm multilayer at a saturation fluence of ∼1 dpa (2 × 1014 ions/cm2). Direct observations of frequent loop absorption by layer interfaces suggest that these interfaces are efficient defect sinks. Ag/Ni 5 nm multilayer showed a superior morphological stability against radiation compared to Ag/Ni 50 nm film.
    Philosophical Magazine 07/2013; 93(26):3547-3562. · 1.60 Impact Factor
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    ANS summer meeting 2013; 06/2013
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    ABSTRACT: In order to study irradiation damage and inert gas bubble formation and growth behaviors, and to provide results and insights useful towards the validation of a multi-scale simulation approach based on a newly developed Xe–Mo inter-atomic potential, in situ Transmission Electron Microscopy (TEM) studies of Xe implantations in pure single crystal Molybdenum (Mo) have been conducted. 300 keV and 400 keV Xe+ ion beams were used to implant Xe in pre-thinned TEM Mo specimens. The irradiations were conducted at 300 °C and 600 °C to ion fluence up to 4 × 1016 ions/cm2.In situ TEM characterization allows detailed behaviors of defect clusters to be observed and is very useful in illustrating defect interaction mechanisms and processes. Dislocation loops were found to form at relatively low irradiation fluence levels. The characterization results showed that the free surfaces, formed in the process of producing pre-thinned specimens, play an important role in influencing the behaviors of dislocation loops. Similar characterizations were conducted at high fluence levels where Xe gas bubbles can be clearly observed. Xe gas bubbles were observed to form by a multi-atom nucleation process and they were immobile throughout the irradiation process at both temperatures. Measurements on both the number density and the size of dislocation loops and gas bubbles were taken. The results and implications of the measurements are discussed in this paper.
    Journal of Nuclear Materials 06/2013; 437(s 1–3):240–249. · 2.02 Impact Factor
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    ABSTRACT: In situ self-ion irradiations (150 keV W+) have been carried out on W and W–5Re at 500 °C, with doses ranging from 1016 to 1018 W+m−2 (∼1.0 dpa). Early damage formation (1016W+m−2) was observed in both materials. Black–white contrast experiments and image simulations using the TEMACI software suggested that vacancy loops were formed within individual cascades, and thus, the loop nucleation mechanism is likely to be ‘cascade collapse’. Dynamic observations showed the nucleation and growth of interstitial loops at higher doses, and that elastic loop interactions may involve changes in loop Burgers vector. Elastic interactions may also promote loop reactions such as absorption or coalescence or loop string formation. Loops in both W and W–5Re remained stable after annealing at 500 °C. One-dimensional hopping of loops (b = 1/2 111>) was only seen in W. At the final dose (1018W+m−2), a slightly denser damage microstructure was seen in W–5Re. Both materials had about 3–4 × 1015 loops m−2. Detailed post-irradiation analyses were carried out for loops of size 4 nm. Both b = 1/2 111 (∼75%) and b = 100> (∼25%) loops were present. Inside–outside contrast experiments were performed under safe orientations to determine the nature of loops. The interstitial-to-vacancy loop ratio turned out close to unity for 1/2 111 loops in W, and for both 1/2 111 and 100 loops in W–5Re. However, interstitial loops were dominant for 100 loops in W. Re seemed to restrict loop mobility, leading to a smaller average loop size and a higher number density in the W-Re alloy.
    Philosophical Magazine A 05/2013; 93(14):1715-1738.
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    ABSTRACT: Significant microstructural damage, in the form of defect clusters, typically occurs in metals subjected to heavy ion irradiation. High angle grain boundaries (GBs) have long been postu-lated as sinks for defect clusters, like dislocation loops. Here, we provide direct evidence, via in situ Kr ion irradiation within a transmission electron microscope, that high angle GBs in nanocrystalline (NC) Ni, with an average grain size of ~55 nm, can effectively absorb irradia-tion-induced dislocation loops and segments. These high angle GBs significantly reduce the density and size of irradiation-induced defect clusters in NC Ni compared to their bulk coun-terparts, and thus NC Ni achieves significant enhancement of irradiation tolerance.
    04/2013; 44A:1966-1974.
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    ABSTRACT: In this study, we report direct observations on heavy ion (Kr2+) irradiation induced changes in microstructures of pure Zr at different temperatures using intermediate voltage electron microscopy. Thin TEM foils were irradiated with 1 MeV Kr2+ ions. Materials have been irradiated to different damage levels ranging from 0.008 dpa to 1 dpa at different temperatures ranging from 300 °C to 500 °C. We particularly concentrate on yield of small defects directly occurring from cascade collapse at very low doses, and their evolution as the dose increases. In situ observation of growth and evolution of these small defects into complex defect structures at high dose has been carried out. Irradiation of materials at different temperatures provided an opportunity to investigate the temperature dependence of defect accumulation in Zr during irradiation. The differences in defect structures, defect densities, and therefore dynamic growth have been discussed in detail as a function of irradiation parameters (dose, temperature). Interaction of irradiation induced defects with existing microstructure and other defects is discussed.
    Journal of Nuclear Materials 02/2013; 433(1-3):138. · 2.02 Impact Factor

Publication Stats

2k Citations
207.95 Total Impact Points

Institutions

  • 1993–2014
    • University of Oxford
      • Department of Materials
      Oxford, England, United Kingdom
  • 1982–2014
    • Argonne National Laboratory
      • • Division of Materials Science
      • • Center for Electron Microscopy
      Lemont, Illinois, United States
  • 2013
    • Texas A&M University
      • Department of Mechanical Engineering
      College Station, TX, United States
  • 1998
    • University of Vienna
      Wien, Vienna, Austria
  • 1991
    • Iowa State University
      Ames, Iowa, United States
    • University of Illinois, Urbana-Champaign
      • Department of Materials Science and Engineering
      Urbana, IL, United States