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논문 기본 정보
- 자료유형
- 학술저널
- 저자정보
- 발행연도
- 2020.1
- 수록면
- 708 - 720 (13page)
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Core-wide temperature distribution in sodium-cooled fast reactor plays a key role in its decay heat removal process, however the prediction for temperature distribution is quite complex due to the conjugate heat transfer between the assembly flow and the inter-wrapper flow. Hybrid medium model has been proposed for conjugate heat transfer modeling in the core. The core is modeled with a Realistic modeled inter-wrapper flow and hybrid medium modeled assembly flow. To validate present model, simulations for a three-assembly model were performed with Realistic modeling, traditional porous medium model and hybrid medium model, respectively. The influences of Uniform/Non-Uniform power distribution among assemblies and the Peclet number within the assembly flow have been considered. Compared to traditional porous medium model, present model shows a better agreement with in Realistic modeling prediction of the temperature distribution and the radial heat transfer between the inter-wrapper flow and the assembly flow.
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참고문헌 (21)
참고문헌 신청
D. Tenchine / 2010 / Some thermal hydraulic challenges in sodium cooled fast reactors / Nucl. Eng. Des 240:1195~1217
N. Yue / 2018 / The development and validation of the inter-wrapper flow model in sodiumcooled fast reactors / Prog. Nucl. Energy 108:54~65
D. Weinberg / 1995 / Transient neptun experiments on passive decay heat removal / Proc. 3rd JSME/ASME Joint Int. Conf. On Nucl. Eng. :519~524
H. Kamide / 2001 / Investigation of core thermohydraulics in fast reactorsdinterwrapper flow during natural circulation / Nucl. Technol 133:77~91
M. Wang / 2017 / Multiscale thermal hydraulic study under the inadvertent safety injection system operation scenario of typical pressurized water reactor / Sci. Technol. Nucl. Install 2017
N. Yue / 2018 / The development and validation of the inter-wrapper flow model in sodiumcooled fast reactors / Prog. Nucl. Energy 108:54~65
D. Weinberg / 1995 / Transient neptun experiments on passive decay heat removal / Proc. 3rd JSME/ASME Joint Int. Conf. On Nucl. Eng. :519~524
H. Kamide / 2001 / Investigation of core thermohydraulics in fast reactorsdinterwrapper flow during natural circulation / Nucl. Technol 133:77~91
M. Wang / 2017 / Multiscale thermal hydraulic study under the inadvertent safety injection system operation scenario of typical pressurized water reactor / Sci. Technol. Nucl. Install 2017
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