Microstructural development in cubic silicon carbide during irradiation at elevated temperatures

Katoh, Yutai; Hashimoto, Naoyuki; Kondo, Sosuke; Snead, Lance Lewis; Kohyama, Akira

doi:10.1016/j.jnucmat.2006.02.007

Cited by 162 publications

(97 citation statements)

References 60 publications

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“…2,19,20 In the present study, quantitative results of their size and bulk number density showed a good agreement with the previous reports. The loops were identified or speculated to be of interstitial type in Ref.…”

Section: Discussionsupporting

confidence: 93%

“…This is supported by the reports showing void formation in SiC at 4.9 dpa, 1050˚C in a neutron study 20 and 10 dpa, 1000˚C in a self-ion study. 19 Note that the population of loops and voids were reported to be nearly saturated by 1 dpa at the irradiation temperature below 1460˚C. 20 In a condition such that most vacancies are no longer isolated, recombination terms can be ignored and the sink strength depending strongly on the population of vacancy clusters may be assumed constant.…”

Section: Discussionmentioning

confidence: 99%

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Analysis of grain boundary sinks and interstitial diffusion in neutron-irradiated SiC

2011

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The widths of interstitial loop denuded zone (DZ) along grain boundaries were examined for 3C-SiC irradiated at 1010-1380˚C by transmission electron microscopy in an effort to obtain the activation energy of interstitial migration. DZ widths as small as 17 nm were observed below 1130˚C, indicating that a substantial population of "TEM invisible" voids of diameter <0.7 significantly contribute to interstitial annihilation. Using the obtained loop DZ width and the matrix sink strength including the invisible voids, the activation energy of interstitial diffusion was determined to be 1.5 eV for the slower moving Si interstitial of SiC by application of simple reaction-diffusion equations.3

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Section: Discussionsupporting

confidence: 93%

Section: Discussionmentioning

confidence: 99%

Analysis of grain boundary sinks and interstitial diffusion in neutron-irradiated SiC

2011

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“…The microstructural changes in 3C-SiC under neutron and self ion irradiation have originally been summarised by Katoh et al [98] and updated by Snead et al [49] in Figure 7 into three overlapping regimes. At low temperatures and low irradiation fluences the main defects are point defects (called black spot defects due to their appearances in weak beam dark field TEM images) and small interstitial clusters in various configurations.…”

Section: Bombardment-induced Defect Typesmentioning

confidence: 99%

Diffusion of fission products and radiation damage in SiC

Malherbe

2013

J. Phys. D: Appl. Phys.

100

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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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“…19) Fig reactors and ion accelerators. While these experiments have elucidated the species and sizes of defect clusters under various irradiation conditions, 25,26) the formation mechanism of defect clusters remains poorly understood. In the present study, the kinetics of defect cluster formation in ¢-SiC were numerically evaluated through a multiscale modeling approach, focusing on the nucleation and growth processes of SIA clusters.…”

Section: Multiscale Modeling Of Microstructural Change In Sic During mentioning

confidence: 99%

Integrated Material System Modeling of Fusion Blanket

Sagara

Nygren

Miyamoto³

et al. 2013

Mater. Trans.

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The blanket of fusion reactors is a multifunctional system that breeds tritium, harvests heat from the burning plasma, and protects the other components and the environment. The common task in the US-J TITAN (Tritium, Irradiation and Thermofluid for America and Nippon) project identifies cross-linked considerations in the blanket system modeling on the basis of the material research in each task. In this paper, we review the main outputs of this task: elucidation of the effect of helium on deuterium retention in the tungsten wall, analysis of tritium transfer in a Pb Li liquid breeder, experimental and computational studies on the effects of radiation damage on hydrogen trapping, and modeling of the tritium barrier in a double-tube heat exchanger.

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Microstructural development in cubic silicon carbide during irradiation at elevated temperatures

Cited by 162 publications

References 60 publications

Analysis of grain boundary sinks and interstitial diffusion in neutron-irradiated SiC

Analysis of grain boundary sinks and interstitial diffusion in neutron-irradiated SiC

Diffusion of fission products and radiation damage in SiC

Integrated Material System Modeling of Fusion Blanket

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