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논문 기본 정보
- 자료유형
- 학술저널
- 저자정보
- 발행연도
- 2016.6
- 수록면
- 660 - 672 (13page)
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In Part I of this paper, the two-temperature homogenized model for the fully ceramicmicroencapsulated fuel, in which tristructural isotropic particles are randomly dispersed in afine lattice stochastic structure, was discussed. In this model, the fuel-kernel and siliconcarbide matrix temperatures are distinguished. Moreover, the obtained temperature profilesare more realistic than those obtained using other models. Using the temperature-dependentthermal conductivities of uranium nitride and the silicon carbide matrix, temperaturedependenthomogenized parameters were obtained. In Part II of the paper, coupled withthe COREDAX code, a reactor core loaded by fully ceramic microencapsulated fuel in whichtristructural isotropic particles are randomly dispersed in the fine lattice stochastic structureis analyzed via a two-temperature homogenized model at steady and transient states. Theresults are compared with those from harmonic- and volumetric-average thermalconductivity models; i.e., we compare keff eigenvalues, power distributions, and temperatureprofiles in the hottest single channel at a steady state. At transient states, we compare totalpower, average energy deposition, and maximumtemperatures in the hottest single channelobtained by the differentthermal analysismodels. The different thermal analysis models andthe availability of fuel-kernel temperatures in the two-temperature homogenized model forDoppler temperature feedback lead to significant differences
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참고문헌 (15)
참고문헌 신청
이윤희 / 2016 / Steady- and Transient-State Analyses of Fully Ceramic Microencapsulated Fuel with Randomly Dispersed Tristructural Isotropic Particles via Two-Temperature Homogenized Model-II: Applications by Coupling with COREDAX / Nuclear Engineering and Technology 48(3):660~672
H. Yu / 2006 / Comparison of Monte Carlo simulation models for randomly distributed particle fuels in VHTR Fuel Element / Trans. Am. Nucl. Soc 95:719~721
B. Cho / 2015 / Theory Manual for the Rectangular Three-Dimensional Diffusion Nodal Code COREDAX-2Version 1.5, KINS/HR-1356 Rev. 1, NURAPT-2015-02 / Korea Advanced Institute of Science and Technology/Korea Institute of Nuclear Safety
S.B. Ross / 1988 / Thermal conductivity correlation for uranium nitride fuel between 10 and 1923 K / J. Nucl. Mater 151:313~317
IAEA / 2006 / Thermophysical Properties Database of Materials for Light Water Reactor and Heavy Water Reactors / IAEATECDOC-1496
H. Yu / 2006 / Comparison of Monte Carlo simulation models for randomly distributed particle fuels in VHTR Fuel Element / Trans. Am. Nucl. Soc 95:719~721
B. Cho / 2015 / Theory Manual for the Rectangular Three-Dimensional Diffusion Nodal Code COREDAX-2Version 1.5, KINS/HR-1356 Rev. 1, NURAPT-2015-02 / Korea Advanced Institute of Science and Technology/Korea Institute of Nuclear Safety
S.B. Ross / 1988 / Thermal conductivity correlation for uranium nitride fuel between 10 and 1923 K / J. Nucl. Mater 151:313~317
IAEA / 2006 / Thermophysical Properties Database of Materials for Light Water Reactor and Heavy Water Reactors / IAEATECDOC-1496
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