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- 저자 : A.G. CLASS (검색결과 2 건)
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CFD ANALYSIS OF HEAVY LIQUID METAL FLOW IN THE CORE OF THE HELIOS LOOP
A. BATTA,JAE HYUN CHO,A.G. CLASS,IL SOON HWANG 한국원자력학회 2010 Nuclear Engineering and Technology Vol.42 No.6
Lead-alloys are very attractive nuclear coolants due to their thermo-hydraulic, chemical, and neutronic properties. Byutilizing the HELIOS (Heavy Eutectic liquid metal Loop for Integral test of Operability and Safety of PEACER2) facility, athermal hydraulic benchmarking study has been conducted for the prediction of pressure loss in lead-alloy cooled advancednuclear energy systems (LACANES). The loop has several complex components that cannot be readily characterized withavailable pressure loss correlations. Among these components is the core, composed of a vessel, a barrel, heaters separatedby complex spacers, and the plenum. Due to the complex shape of the core, its pressure loss is comparable to that of the restof the loop. Detailed CFD simulations employing different CFD codes are used to determine the pressure loss, and it is foundthat the spacers contribute to nearly 90 percent of the total pressure loss. In the system codes, spacers are usually accountedfor; however, due to the lack of correlations for the exact spacer geometry, the accuracy of models relies strongly on assumptionsused for modeling spacers. CFD can be used to determine an appropriate correlation. However, application of CFD alsorequires careful choice of turbulence models and numerical meshes, which are selected based on extensive experience withliquid metal flow simulations for the KALLA lab. In this paper consistent results of CFX and Star-CD are obtained andcompared to measured data. Measured data of the pressure loss of the core are obtained with a differential pressure transducerlocated between the core inlet and outlet at a flow rate of 13.57kg/s
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CFD ANALYSIS OF HEAVY LIQUID METAL FLOW IN THE CORE OF THE HELIOS LOOP
Batta, A.,Cho, Jae-Hyun,Class, A.G.,Hwang, Il-Soon Korean Nuclear Society 2010 Nuclear Engineering and Technology Vol.42 No.6
Lead-alloys are very attractive nuclear coolants due to their thermo-hydraulic, chemical, and neutronic properties. By utilizing the HELIOS (Heavy Eutectic liquid metal Loop for Integral test of Operability and Safety of PEACER$^2$) facility, a thermal hydraulic benchmarking study has been conducted for the prediction of pressure loss in lead-alloy cooled advanced nuclear energy systems (LACANES). The loop has several complex components that cannot be readily characterized with available pressure loss correlations. Among these components is the core, composed of a vessel, a barrel, heaters separated by complex spacers, and the plenum. Due to the complex shape of the core, its pressure loss is comparable to that of the rest of the loop. Detailed CFD simulations employing different CFD codes are used to determine the pressure loss, and it is found that the spacers contribute to nearly 90 percent of the total pressure loss. In the system codes, spacers are usually accounted for; however, due to the lack of correlations for the exact spacer geometry, the accuracy of models relies strongly on assumptions used for modeling spacers. CFD can be used to determine an appropriate correlation. However, application of CFD also requires careful choice of turbulence models and numerical meshes, which are selected based on extensive experience with liquid metal flow simulations for the KALLA lab. In this paper consistent results of CFX and Star-CD are obtained and compared to measured data. Measured data of the pressure loss of the core are obtained with a differential pressure transducer located between the core inlet and outlet at a flow rate of 13.57kg/s.
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