Recession of Silicon Carbide in Steam under Nuclear Plant LOCA Conditions up to 1400 °C

Markham, Greg; Hall, R.C.; Feinroth, H.

doi:10.1002/9781118217535.ch8

Cited by 7 publications

(6 citation statements)

References 8 publications

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“…Since the oxidation reaction is not gas transport limited under these conditions, the parabolic rate constant depends solely on the steam partial pressure. Since many of the previous studies focused on combustion turbine applications, experimental results on the oxidation of SiC in pure steam environments in this temperature range are scarce . Some discussion along with experimental data are available in literature .…”

Section: Analysis–monolithic Cvd Sicmentioning

confidence: 99%

Silicon Carbide Oxidation in Steam up to 2 MPa

et al. 2014

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Growth and microstructure of a protective or nonprotective SiO2 scale and the subsequent volatilization of scale formed on high‐purity chemical vapor deposited (CVD) SiC and nuclear‐grade SiC/SiC composites have been studied during high‐temperature 100% steam exposure. The environmental parameters of interest were temperature from 1200°C to 1700°C, pressure of 0.1 to 2 MPa and flow velocities of 0.23 to 145 cm/s. Scale microstructure was characterized via electron microscopy and X‐ray diffractometry. The Arrhenius dependence of the parabolic oxidation and linear volatilization rate constants were determined. The linear volatilization rate exhibited a strong dependence on steam partial pressure with a weaker dependence on flow velocity. At high steam pressures, the oxide scale developed substantial porosity, which significantly accelerated material recession. The dominant oxide phase for the conditions studied was cristobalite. The oxidation behavior of SiC/SiC composite was strongly dependent on the state of the surface, specifically whether steam could find easy entry into the material via surface‐exposed interface layers. For the case where these as‐machined interfaces were surface coated with matrix CVD SiC, composite recession was found to be essentially that of high‐purity CVD SiC.

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Section: Analysis–monolithic Cvd Sicmentioning

confidence: 99%

Silicon Carbide Oxidation in Steam up to 2 MPa

et al. 2014

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“…SiC ceramics show outstanding oxidation resistance and a low hydrogen liberation rate in hot steam, compared with the current Zr alloys. Thus, SiC ceramics promise larger safety margins under severe accident conditions of PWRs [15][16][17]. Moreover, their high temperature strength and stability under high neutron doses also provide high burn-up capability [11].…”

Section: Introductionmentioning

confidence: 99%

Effect of dissolved hydrogen on the corrosion behavior of chemically vapor deposited SiC in a simulated pressurized water reactor environment

et al. 2015

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“…Another way to mitigate against severe accidents is to develop enhanced strength and ductility ATF cladding, as this would alleviate the severity of reactivity-initiated accidents, and oxidation-resistant ATF cladding in the case of a loss-of-coolant accident (LOCA) [3]. Zr alloys, which have low neutron absorption cross-sections and a good neutron irradiation resistance, are currently used in nuclear fuel cladding for commercial LWRs [4]. However, in accident scenarios, Zr alloys react with hot The as-fabricated thermal conductivity of UO 2 -BeO fuel is calculated from the Hasselman and Johnson model, as shown below [10].…”

Section: Introductionmentioning

confidence: 99%

Thermophysical and Mechanical Analyses of UO2-36.4vol % BeO Fuel Pellets with Zircaloy, SiC, and FeCrAl Claddings

Zhou

2018

Metals

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Abstract:The thermophysical performance and solid mechanics behavior of UO 2 -36.4vol % BeO fuel pellets cladded with Zircaloy, SiC, and FeCrAl, and Zircaloy cladding materials coated with SiC and FeCrAl, are investigated based on simulation results obtained by the CAMPUS code. In addition, the effect of coating thickness (0.5, 1 and 1.5 mm) on fuel performance and mechanical interaction is discussed. The modeling results show that Zircaloy claddings are more effective in decreasing fuel centerline temperature and fission gas release than other kinds of cladding material because of the smaller gap between cladding and fuel at the same burnup. SiC claddings and SiC-coated Zircaloy claddings possess smaller plenum pressure than other kinds of cladding. SiC claddings contribute more to fuel radial displacement but less to fuel axial displacement. FeCrAl claddings exhibit very different radial and axial displacements in different axial positions. FeCrAl-coated Zircaloy claddings have a lower fuel centerline temperature than Zircaloy claddings at burnup below 850 MWh/kg U, but a higher fuel centerline temperature at higher burnup. The gap between FeCrAl-coated Zircaloy claddings and fuel pellets closes earlier than that of Zircaloy claddings. SiC-coated claddings increase fuel radial and axial displacements, and cladding axial displacements of inner and outer cladding surfaces.

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Recession of Silicon Carbide in Steam under Nuclear Plant LOCA Conditions up to 1400 °C

Cited by 7 publications

References 8 publications

Silicon Carbide Oxidation in Steam up to 2 MPa

Silicon Carbide Oxidation in Steam up to 2 MPa

Effect of dissolved hydrogen on the corrosion behavior of chemically vapor deposited SiC in a simulated pressurized water reactor environment

Thermophysical and Mechanical Analyses of UO2-36.4vol % BeO Fuel Pellets with Zircaloy, SiC, and FeCrAl Claddings

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