Ag diffusion in cubic silicon carbide

Shrader, David; Khalil, Sarah; Gerczak, Tyler J.; Allen, Todd R.; Heim, Andrew J.; Szlufarska, Izabela; Morgan, Dane

doi:10.1016/j.jnucmat.2010.10.088

Cited by 99 publications

(127 citation statements)

References 37 publications

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“…49 Therefore, the migration barrier of 3.2eV for V Si 1-reported by M.Bockstedte et al 15 was adopted. 2 A lower value of about 0.8 eV 43,50 has been reported for the migration barrier of Si I . However, based on the dependence of interstitial loop denuded zone (DZ) width on temperature along grain boundaries in 3C-SiC, 46 the activation energy for migration of Si interstitials has been determined to be about 1.5eV.…”

Section: Okmc Simulation For Long-term Defect Evolutionmentioning

confidence: 94%

“…Most of these parameters come from a recent ab-initio calculation by D. Shraders et al, 43 while a few parameters are adopted from other works 44,46 for the following reasons: 1…”

Section: Okmc Simulation For Long-term Defect Evolutionmentioning

confidence: 99%

“…43 has proposed a lower migration barrier of 2.7eV for V Si in n-doping 3C-SiC, which corresponds to a charge state of -2. However, the experimental observed V Si possess a -1 charge state, due to its high-spin configuration (S=3/2), and has been unambiguously identified the T1 center, [47][48][49] which can be detected in n-type and p-type 3C-SiC irradiated by electrons and protons.…”

Section: Okmc Simulation For Long-term Defect Evolutionmentioning

confidence: 99%

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Modeling of long-term defect evolution in heavy-ion irradiated 3C-SiC: Mechanism for thermal annealing and influences of spatial correlation

Guo

Martín-Bragado

et al. 2014

Journal of Applied Physics

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Based on the parameters from published ab-inito theoretical and experimental studies, and combining Molecular Dynamics (MD) and kinetic Monte Carlo (KMC) simulations, a framework of multi-scale modeling is developed to investigate the long-term evolution of displacement damage induced by heavy-ion irradiation in cubic silicon carbide. The isochronal annealing after heavy ion irradiation is simulated, and the annealing behaviors of total interstitials are found consistent with previous experiments. Two annealing stages below 600K and one stage above 900K are identified. The mechanisms for those recovery stages are interpreted by the evolution of defects. The influence of the spatial correlation in primary damage on defect recovery has been studied and found insignificant when the damage dose is high enough, which sheds light on the applicability of approaches with mean-field approximation to the long-term evolution of damage by heavy ions in SiC.

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Section: Okmc Simulation For Long-term Defect Evolutionmentioning

confidence: 94%

“…Most of these parameters come from a recent ab-initio calculation by D. Shraders et al, 43 while a few parameters are adopted from other works 44,46 for the following reasons: 1…”

Section: Okmc Simulation For Long-term Defect Evolutionmentioning

confidence: 99%

Section: Okmc Simulation For Long-term Defect Evolutionmentioning

confidence: 99%

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Modeling of long-term defect evolution in heavy-ion irradiated 3C-SiC: Mechanism for thermal annealing and influences of spatial correlation

Guo

Martín-Bragado

et al. 2014

Journal of Applied Physics

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“…Even from the early studies (e.g. [254], and especially from the newer silver implantation studies in single crystal SiC at elevated temperatures where the SiC remains crystalline [100 -105, 270, 283] and ab initio simulation studies [277,278,282] it is clear that the volume diffusion coefficient of silver in SiC is too low to account for the silver release data. In one of the early studies [254] it was proposed that the release was associated with the migration of silver through grain boundaries of the coated polycrystalline SiC enhanced by traces of free silicon.…”

Section: Silver Diffusionmentioning

confidence: 99%

Diffusion of fission products and radiation damage in SiC

Malherbe

2013

J. Phys. D: Appl. Phys.

109

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A major problem with most of the present nuclear reactors is their safety in terms of the release of radioactivity into the environment during accidents. In some of the future nuclear reactor designs, i.e. Generation IV reactors, the fuel is in the form of coated spherical particles, i.e. TRISO (acronym for Triple Coated Isotropic) particles. The main function of these coating layers is to act as diffusion barriers for radioactive fission products, thereby keeping these fission products within the fuel particles, even under accident conditions. The most important coating layer is composed of polycrystalline 3C-SiC. This paper reviews the diffusion of the important fission products (silver, caesium, iodine and strontium) in SiC. Because radiation damage can induce and enhance diffusion, the paper also briefly reviews damage created by energetic neutrons and ions at elevated temperatures, i.e. the temperatures at which the modern reactors will operate, and the annealing of the damage. The interaction between SiC and some fission products (such as Pd and I) is also briefly discussed. As shown, one of the key advantages of SiC is its radiation hardness at elevated temperatures, i.e. SiC is not amorphized by neutron or bombardment at substrate temperatures above 350°C. Based on the diffusion coefficients of the fission products considered, the review shows that at the normal operating temperatures of these new reactors (i.e. less than 950°C) the SiC coating layer is a good diffusion barrier for these fission products. However, at higher temperatures the design of the coated particles needs to be adapted, possibly by adding a thin layer of ZrC.

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“…Grain boundary diffusion has been suggested and discussed as a possible transport mechanism in neutron irradiated TRISO coated particles as early as 1975 by Nabielek et al, [5]. More recent out-of-pile experimental on SiC and computational work also indicated that Ag transports via grain boundary diffusion under their specific experimental conditions [6][7][8][9][10]. Experimental evidence by Lopez-Honorato, et al, [8] using scanning electron microscopy; indicated that silver accumulated at grain boundaries in unirradiated SiC layers of a simulated TRISO coated particle experiment.…”

Section: Introductionmentioning

confidence: 99%

Identification of silver and palladium in irradiated TRISO coated particles of the AGR-1 experiment

Rooyen

Lillo

2014

Journal of Nuclear Materials

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Evidence of the release of certain metallic fission products through intact tristructural isotropic (TRISO) particles has been seen for decades around the world, as well as in the recent AGR-1 experiment at the Idaho National Laboratory (INL). However, understanding the basic mechanism of transport is still lacking. This understanding is important because the TRISO coating is part of the high temperature gas-cooled reactor functional containment and critical for the safety strategy for licensing purposes.Our approach to identify fission products in irradiated AGR-1 TRISO fuel using scanning transmission electron microscopy (STEM), Electron Energy-Loss Spectroscopy (EELS) and Energy Filtered TEM (EFTEM), has led to first-of-a-kind data at the nano-scale indicating the presence of silver at triple-points and grain boundaries of the SiC layer in the TRISO particle. Cadmium was also found in the triple junctions. In this initial study, the silver was only identified in SiC grain boundaries and triple points on the edge of the SiC-IPyC interface up to a depth of approximately 0.5 μm.Palladium was identified as the main constituent of micron-sized precipitates present at the SiC grain boundaries. Additionally spherical nano-sized palladium rich precipitates were found inside the SiC grains. No silver was found in the center of the micron-sized fission product precipitates using these techniques, although silver was found on the outer edge of one of the Pd-U-Si containing precipitates which was facing the IPyC layer. Only Pd-U containing precipitates were identified in the IPyC layer and no silver was identified in the IPyC layer.The identification of silver alongside the SiC grain boundaries and the findings of Pd inside the SiC grains and alongside SiC grain boundaries provide important information needed to understand silver and palladium transport in TRISO fuel, which has been the topic of international research for the past forty years. The findings reported in this paper may support the postulations of recent research that Ag transport may be driven by grain boundary diffusion. However, more work is needed to fully understand the transport mechanisms. Additionally, the usefulness of the advanced electron microscopic techniques for TRISO coated particle research is demonstrated in this paper.

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Ag diffusion in cubic silicon carbide

Cited by 99 publications

References 37 publications

Modeling of long-term defect evolution in heavy-ion irradiated 3C-SiC: Mechanism for thermal annealing and influences of spatial correlation

Modeling of long-term defect evolution in heavy-ion irradiated 3C-SiC: Mechanism for thermal annealing and influences of spatial correlation

Diffusion of fission products and radiation damage in SiC

Identification of silver and palladium in irradiated TRISO coated particles of the AGR-1 experiment

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