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논문 기본 정보
- 자료유형
- 학술저널
- 저자정보
- 발행연도
- 2019.1
- 수록면
- 73 - 83 (11page)
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It is important to clarify the mechanisms of molten-fuel-pool sloshing behavior that might be encounteredduring a core disruptive accident of sodium-cooled fast reactors. In this study, motivated byacquiring some evidence for understanding the characteristics of this behavior at more realistic conditions,a number of experiments are newly performed by injecting nitrogen gas into a water pool with theaccumulation of solid particles. To achieve comprehensive understanding, various parameters includingparticle bed height, particle size, density, shape, gas pressure along with the gas-injection duration, wereemployed. It is found that due to the different interaction mechanisms between solid particles and thegas bubble injected, three kinds of regimes, termed respectively as the bubble-impulsion dominantregime, the transitional regime and the bed-inertia dominant regime, could be identified. The performedanalyses also suggest that under present conditions, all our experimental parameters employed can havenoticeable impact on the regime transition and resultant sloshing intensity (e.g. maximum elevation ofwater level at pool peripheries). Knowledge and fundamental data from this work will be used for thefuture verifications of fast reactor severe accident codes in China
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참고문헌 (16)
참고문헌 신청
Songbai Cheng / 2013 / Characteristics of Self-Leveling Behavior of Debris Beds in a Series of Experiments / Nuclear Engineering and Technology 45(3):323~334
T. Suzuki / 2014 / A scenario of core disruptive accident for Japan sodium-cooled fast reactor to achieve in-vessel retention / J. Nucl. Sci. Technol 51(4):493~513
T. G. Theofanous / 1986 / An assessment of Clinch River breeder reactor core disruptive accident energetics / Nucl. Technol 93(3):215~228
W. Maschek / 1992 / Investigations of sloshing fluid motions in pools related to recriticalities in liquid-metal fast breeder reactor core meltdown accidents / Nucl. Technol 98(1):27~43
H. Yamano / 2009 / Transient heat transfer characteristics between molten fuel and steel with steel boiling in the CABRI-TPA2 test / Nucl. Technol 165(2):145~165
T. Suzuki / 2014 / A scenario of core disruptive accident for Japan sodium-cooled fast reactor to achieve in-vessel retention / J. Nucl. Sci. Technol 51(4):493~513
T. G. Theofanous / 1986 / An assessment of Clinch River breeder reactor core disruptive accident energetics / Nucl. Technol 93(3):215~228
W. Maschek / 1992 / Investigations of sloshing fluid motions in pools related to recriticalities in liquid-metal fast breeder reactor core meltdown accidents / Nucl. Technol 98(1):27~43
H. Yamano / 2009 / Transient heat transfer characteristics between molten fuel and steel with steel boiling in the CABRI-TPA2 test / Nucl. Technol 165(2):145~165
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