Microstructural defects in SiC neutron irradiated at very high temperatures

Kondo, Sosuke; Katoh, Yutai; Snead, Lance Lewis

doi:10.1016/j.jnucmat.2008.08.013

Cited by 68 publications

(30 citation statements)

References 38 publications

(43 reference statements)

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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%

“…The absence of vacancy-type loops implies the existence of tiny three-dimensional clusters of vacancies for the temperatures studied here, though the number density of "TEM visible" voids was zero at 1010˚C or very limited at 1050˚C. 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.…”

Section: Discussionsupporting

confidence: 80%

“…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. Then, the steadystate reaction-diffusion equation near the one-dimensional sinks such as grain boundary is conventionally given as: are assumed to be continuously unsaturated with the migrating point defects (in other words treated as a perfect sink) in accordance with generally accepted boundary condition.…”

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: Discussionsupporting

confidence: 80%

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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“…From Figure 7 it can be seen that vacancy clusters in the form of voids appear at relatively high fluences and at high temperatures where vacancies are sufficiently mobile [157,159,173,174]. The voids are faceted and appeared to be tetrahedrally bounded by {111} planes.…”

Section: Bombardment-induced Defect Typesmentioning

confidence: 99%

Diffusion of fission products and radiation damage in SiC

Malherbe

2013

J. Phys. D: Appl. Phys.

105

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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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“…In the temperature regime between ~200-800ºC, the swelling is associated with the accumulation of point defects and/or its clusters [5,7]. The anisotropic distribution of the defects is suspected to be one of the causes for the irradiation creep [8,9]. The linear relationship observed in Fig.…”

Section: Resultsmentioning

confidence: 99%

Development and Testing of Ion-Irradiation under the Applied Stress for Nuclear Ceramics

Kondo

Koyanagi

Hinoki

2011

IOP Conf. Ser.: Mater. Sci. Eng.

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Abstract. Simultaneous effects of the ion-irradiation and applied stress on the dimensional stability of SiC was studied up to 1 dpa at the irradiation temperature of 800ºC. Tensioned surface of the curved SiC strips were irradiated using fixtures developed, and the irradiation creep strain of the irradiated samples was estimated from the irradiated curvatures. The irradiation creep rates were nearly proportional to the swelling rates. Linear relationship between the irradiation creep strain and stress was found below 1 dpa for the stress levels ranging 150-300 MPa.

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Microstructural defects in SiC neutron irradiated at very high temperatures

Cited by 68 publications

References 38 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

Development and Testing of Ion-Irradiation under the Applied Stress for Nuclear Ceramics

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