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논문 기본 정보
- 자료유형
- 학술저널
- 저자정보
- 발행연도
- 2009.1
- 수록면
- 969 - 978 (10page)
이용수
초록· 키워드
오류제보하기
The ASSERT-PV subchannel code developed by AECL has been applied as a design-assist tool to the advanced CANDU®1
reactor fuel bundle. Based primarily on the CANFLEX®2 fuel bundle, several geometry changes (such as element sizes and pitchcircle
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.
목차
등록된 정보가 없습니다.
참고문헌 (7)
참고문헌 신청
D.C. Groeneveld / 1996 / The 1995 Look-up Table for Critical Heat Flux in Tubes / Nuclear Engineering and Design 163 : 1 ~ 23
G.R. Dimmick / 1999 / Full-Scale Water CHF Testing of the CANFLEX Bundle / Proceedings of the 6th Int. Conf. on CANDU Fuel
L.K.H Leung / 2000 / Fluid-to-Fluid Modelling of Critical Heat Flux in 37-Element bundles / Proceedings of the 21st Nuclear Simulation Symposium
L.N. Carlucci / 2004 / Two-Phase Turbulent Mixing and Buoyancy Drift in Rod Bundles / Nuclear Engineering and Design 227 : 65 ~ 84
M.B. Carver / 1990 / Simulation of flow and phase distribution in vertical and horizontal bundles using the ASSERT subchannel code / Nuclear Engineering and Design 122 : 413
G.R. Dimmick / 1999 / Full-Scale Water CHF Testing of the CANFLEX Bundle / Proceedings of the 6th Int. Conf. on CANDU Fuel
L.K.H Leung / 2000 / Fluid-to-Fluid Modelling of Critical Heat Flux in 37-Element bundles / Proceedings of the 21st Nuclear Simulation Symposium
L.N. Carlucci / 2004 / Two-Phase Turbulent Mixing and Buoyancy Drift in Rod Bundles / Nuclear Engineering and Design 227 : 65 ~ 84
M.B. Carver / 1990 / Simulation of flow and phase distribution in vertical and horizontal bundles using the ASSERT subchannel code / Nuclear Engineering and Design 122 : 413
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