Ultra-high temperature steam corrosion of complex silicates for nuclear applications: A computational study

Rashkeev, Sergey N.; Glazoff, Michael V.; Tokuhiro, Akira

doi:10.1016/j.jnucmat.2013.09.021

Cited by 10 publications

(6 citation statements)

References 26 publications

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“…The SiO 2 scale produced by the oxidation of SiC was around 1 µm thick with single measurement values between 0.5 and 1.6 µm for the SiC-03 sample and 2.7-3.9 µm for the SiC-04 sample. Taking into account the roughly constant gas release rates during the isothermal tests, it could be assumed that the oxide scale thicknesses correspond to the equilibrium value of the paralinear oxidation determined by growth and volatilization of SiO 2 [19]. The oxide thicknesses measured on these samples correspond well to other data published for oxidation of SiC in steam at 1700 • C [13,14,16], which are all in the range of a few micrometers at this temperature.…”

Section: Post-test Examinationssupporting

confidence: 76%

“…The SiC oxidation, especially the SiO 2 volatilization rates are of course dependent on the experimental boundary conditions like gas flow rate, steam partial pressure, and total pressure as, e.g., described experimentally [16] and by modelling [25]. Other authors explicitly mention SiC recession rates of a few µm per hour at 1700 • C for TRISO (TRi-structural ISOtropic fuel) particles [26] or by thermodynamic and firstprinciples atomic-scale calculations [19]. These are lower rates than the ones seen in these tests for larger samples of prototypical cladding tube segments.…”

Section: Discussion Of Resultsmentioning

confidence: 99%

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Oxidation of Silicon Carbide Composites for Nuclear Applications at Very High Temperatures in Steam

et al. 2022

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Single-rod oxidation and quench experiments at very high temperatures in steam atmosphere were conducted with advanced, nuclear grade SiCf/SiC CMC cladding tube segments. A transient experiment was performed until severe local degradation of the sample at maximum temperature of approximately 1845 °C. The degradation was caused by complete consumption of the external CVD-SiC sealcoat, resulting in steam access to the fiber–matrix composite with less corrosion resistance. Approaching these very high temperatures was accompanied by accelerated gas release mainly of H2 and CO2, the formation of surface bubbles and white smoke. Three one-hour isothermal tests at 1700 °C in steam with final water flooding and one three-hour experiment with fast cool-down in Ar atmosphere were run under nominally identical conditions. All isothermally tested samples survived the tests without any macroscopic degradation. The mechanical performance of these quenched clad segments was not significantly affected, while maintaining a high capability to tolerate damages. Despite these harsh exposure conditions, load transfer between SiC fibers and matrix remained efficient, allowing the composites to accommodate deformation.

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Section: Post-test Examinationssupporting

confidence: 76%

Section: Discussion Of Resultsmentioning

confidence: 99%

Oxidation of Silicon Carbide Composites for Nuclear Applications at Very High Temperatures in Steam

et al. 2022

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“…10 Mullite, stabilized zirconia, mullite/stabilized zirconia composite, alumina, yttria-alumina composite, yttrium aluminum garnet, yttrium aluminate perovskite, rare earth silicates, and zirconium silicate are investigated as coating materials to resist oxidation and corrosion of SiC in water vapor containing oxidation environments. [49][50][51][52] Recently, it has been shown that Li 2 O-Al 2 O 3 -SiO 2 -based glaze resists alkali vapor corrosion. 53 These coating materials exhibit limitations despite significant success for oxidation and corrosion resistance.…”

Section: Mahapatramentioning

confidence: 99%

Review of corrosion of refractory in gaseous environment

Mahapatra

2019

Int J Applied Ceramic Tech

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Advancing high temperature technologies pushes the envelope of the service conditions to new extremes. However, corrosion of refractories substantially decreases the service life and causes unscheduled shutdown of manufacturing and processing plants. In this paper, corrosion of refractories in gaseous environments is reviewed.The corrosion mechanisms of alumina refractories, including alumino-silicates and zirconia as oxide, and silicon carbide as non-oxide refractory are emphasized due to their broad array of applications. It is shown that carbonaceous atmosphere and water vapor are detrimental for the oxide refractories. Interaction with water vapor leads to formation of the volatile species and subsequent recession of the refractories.Carbonaceous atmosphere leads to the decomposition of even mullite, which is categorized as neutral refractory due to its excellent thermochemical stability. Oxidizing atmosphere with water vapor dominates the corrosion of non-oxide refractories.Presence of impurities, even parts per million level, in the gas stream such as alkali vapor and halogens can alter the corrosion mechanisms and favor "hot corrosion."

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“…Moreover, spallation of the zircon layer on the local surface was revealed, being related to the weight loss. Even though zircon is one of the most corrosion resistant materials under water conditions [62], it seems that corrosion started at grain boundaries with elimination of silicon species and weakened mechanical integrity of the zircon layer. Ueno et al also reported the same phenomenon during static water corrosion test (1300 °C for 100 hours) for bulk zircon samples [63].…”

Section: Immobilization Of Silicon In Water Corrosionmentioning

confidence: 99%

Development of Self-Healing Zirconium-Silicide Coatings for Improved Performance Zirconium-Alloy Fuel Cladding

Sridharan

Mariani

Bai

et al. 2018

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Zirconium-alloy fuel claddings have been used successfully in Light Water Reactors (LWR) for over four decades. However, under high temperature accident conditions, zirconium-alloys fuel claddings exhibit profuse exothermic oxidation accompanied by release of hydrogen gas due to the reaction with water/steam. Additionally, the ZrO2 layer can undergo monoclinic to tetragonal to cubic phase transformations at high temperatures which can induce stresses and cracking. These events were unfortunately borne out in the Fukushima-Daiichi accident in in Japan in 2011. In reaction to such accident, protective oxidation-resistant coatings for zirconium-alloy fuel claddings has been extensively investigated to enhance safety margins in accidents as well as fuel performance under normal operation conditions. Such surface modification could also beneficially affect fuel rod heat transfer characteristics. Zirconium-silicide, a candidate coating material, is particularly attractive because zirconium-silicide coating is expected to bond strongly to zirconium-alloy substrate. Intermetallic compound phases of zirconium-silicide have high melting points and oxidation of zirconium silicide produces highly corrosion resistant glassy zircon (ZrSiO4) and silica (SiO2) which possessing self-healing qualities. Given the long-term goal of developing such coatings for use with nuclear reactor fuel cladding, this work describes results of oxidation and corrosion behavior of bulk zirconium-silicide and fabrication of zirconium-silicide coatings on zirconium-alloy test flats, tube configurations, and SiC test flats. In addition, boiling heat transfer of these modified surfaces (including ZrSi2 coating) during clad quenching experiments is discussed in detail. Development of Self-Healing Zirconium-Silicide Coatings for Improved Performance Zirconium-Alloy Fuel Cladding DE-NE0008300 Final Report 5 was also confirmed that the theoretical composition coating enhanced oxidation protection and wear resistance. Distinctive oxide layers of ZrSi2 prepared at 1000 °C and 1400 °C in ambient air were subjected to a 3.9 MeV Si 2+ ions irradiation at 305 °C and their radiation responses were characterized and analyzed. Nanocomposite oxides consisting of ZrO2 nanocrystals embedded in amorphous SiO2 matrix formed on ZrSi2 surface after oxidation at 1000 °C. Radiation-induced phase mixing of the oxide phases and amorphization of ZrO2 was observed up to ~ 820 nm in depth, about one-third of the total radiation damaged region (2.5 µm in thickness) as estimated by SRIM calculation. The ion-mixing was facilitated with fine ZrO2 grains (< 25 nm) at low doses (5 to 10 dpa) but not in larger grains (> 30 nm) even at high damage levels (30 to 60 dpa). Qualitative explanation of the ion-mixing of the nanocomposite was proposed with respect to grain size of ZrO2 in SiO2 matrix. Polygonal crystalline ZrSiO4 grains in dual-layered oxide scale on ZrSi2 at 1400 °C were completely amorphized under the ion-irradiation. Given the high corrosion resistance of ZrSiO4 and immobil...

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Ultra-high temperature steam corrosion of complex silicates for nuclear applications: A computational study

Cited by 10 publications

References 26 publications

Oxidation of Silicon Carbide Composites for Nuclear Applications at Very High Temperatures in Steam

Oxidation of Silicon Carbide Composites for Nuclear Applications at Very High Temperatures in Steam

Review of corrosion of refractory in gaseous environment

Development of Self-Healing Zirconium-Silicide Coatings for Improved Performance Zirconium-Alloy Fuel Cladding

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