Development of thermal hydraulic analysis code for nuclear reactors with annular fuels and assessment of the KAIST DNB-type theoretical critical heat flux model = Annular fuel이 장착된 원자로의 열수력 분석 코드 개발 및 KAIST의 DNB 영역의 이론적 임계열유속 모델 평가

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The development of thermal hydraulic analysis code for Gas-Cooled Reactors (GCRs) and for annular fuel and its application to various types of nuclear reactors, and the assessment of the Korea Advanced Institute of Science and Technology (KAIST) Departure from Nucleate Boiling (DNB)-type theoretical Critical Heat Flux (CHF) model for rod bundles with non-uniform axial power shapes were investigated. Thermal hydraulic characteristics of thorium-based fuel assemblies with annular seed pins were analyzed using Thermal-Hydraulic analysis code for Annular Fuel (THAF) combined with Multichannel Analyzer for steady states and Transients in Rod Arrays (MATRA), and compared with those of existing thorium-based assemblies. This study investigates the possibilities of using annular fuel pins in a pressurized water reactor with emphasis on coolant flow distribution and heat transfer fraction in internal and external sub-channels. MATRA and THAF showed good agreements for the pressure drops at the internal sub-channels. Mass fluxes were high in inner sub-channels of the seed pins due to the grid form losses in the outer sub-channels. About 43% of heat generated from the seed pin flowed into the inner sub-channel. The remaining heat flowed into the outer sub-channel. The inner to outer wall heat flux ratio was approximately 1.2. Maximum temperatures of annular seed pins were slightly above 500℃. Minimum DNB Ratios (MDNBRs) of the assemblies loaded with annular seed pins were higher than those of the existing assemblies. Temperatures and enthalpies were higher in the inner sub-channels due to the fact that inter-channel mixing cannot occur in the inner sub-channels. A thermal-hydraulic analysis code for annular fuel-based Liquid Metal Reactors (LMRs) has been developed. About 41% of the heat generated from the fuel pin flowed into the inner sub-channel and the rest into the outer sub-channel. The inner to outer wall heat flux ratio was equal to approximately 1.44. A new 37 an...
Chang, Soon-Heungresearcher장순흥researcher
한국과학기술원 : 원자력및양자공학과,
Issue Date
254260/325007  / 020025317

학위논문(박사) - 한국과학기술원 : 원자력및양자공학과, 2006.2, [ xviii, 124 p. ]


gas-cooled reactor; Annular fuel; critical heat flux; 임계열유속; 가스냉각로; 환형 핵연료

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