Nuclear Engineering and Technology

  • 제41권7호
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  • Pages.969-978
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  • 2009
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  • 1738-5733(pISSN)
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  • 2234-358X(eISSN)
한국원자력학회 (Korean Nuclear Society)
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PREDICTIONS OF CRITICAL HEAT FLUX USING THE ASSERT-PV SUBCHANNEL CODE FOR A CANFLEX VARIANT BUNDLE

  • Onder, Ebru Nihan (Atomic Energy of Canada Limited, Thermalhydraulics Branch, Chalk River Laboratories) ;
  • Leung, Laurence Kim-Hung (Atomic Energy of Canada Limited, Thermalhydraulics Branch, Chalk River Laboratories) ;
  • Rao, Yanfei (Atomic Energy of Canada Limited, Thermalhydraulics Branch, Chalk River Laboratories)
  • 발행 : 2009.09.30
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초록

The ASSERT-PV subchannel code developed by AECL has been applied as a design-assist tool to the advanced $CANDU^{(R)1}$ reactor fuel bundle. Based primarily on the $CANFLEX^{(R)2}$ fuel bundle, several geometry changes (such as element sizes and pitch-circle diameters of various element rings) were examined to optimize the dryout power and pressure-drop performances of the new fuel bundle. An experiment was performed to obtain dryout power measurements for verification of the ASSERT-PV code predictions. It was carried out using an electrically heated, Refrigerant-134a cooled, fuel bundle string simulator. The axial power profile of the simulator was uniform, while the radial power profile of the element rings was varied simulating profiles in bundles with various fuel compositions and burn-ups. Dryout power measurements are predicted closely using the ASSERT-PV code, particularly at low flows and low pressures, but are overpredicted at high flows and high pressures. The majority of data shows that dryout powers are underpredicted at low inlet-fluid temperatures but overpredicted at high inlet-fluid temperatures.

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참고문헌

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