Title of article :
Critical heat fluxes of subcooled water flow boiling in a short vertical tube at high liquid Reynolds number Original Research Article
Author/Authors :
Koichi Hata، نويسنده , , Suguru Masuzaki، نويسنده ,
Issue Information :
روزنامه با شماره پیاپی سال 2010
Pages :
13
From page :
3145
To page :
3157
Abstract :
The steady state critical heat fluxes (CHFs) and the heat transfer of the subcooled water flow boiling for the flow velocities (u = 17.2–42.4 m/s), the inlet subcoolings (ΔTsub,in = 80.9–147.6 K), the inlet pressures (Pin = 812.1–1181.5 kPa) and the exponentially increasing heat input (Q0 exp(t/τ), τ = 8.5 s) are systematically measured by the experimental water loop comprised of a new multi-stage canned-type circulation pump with high pump head. The SUS304 test tube of inner diameter (d = 6 mm), heated length (L = 59.5 mm), L/d = 9.92 and wall thickness (δ = 0.5 mm) with surface roughness (Ra = 3.18 μm) is used in this work. The steady state CHFs of the subcooled water flow boiling for the flow velocities ranging from 17.2 to 42.4 m/s are clarified. The steady state CHFs are compared with the values calculated by our transient CHF correlations against outlet and inlet subcoolings based on the experimental data for the flow velocities ranging from 4.0 to 13.3 m/s. The influence of flow velocity at high liquid Reynolds number on the subcooled flow boiling CHF is investigated in detail and the widely and precisely predictable correlations of the transient CHF correlations against outlet and inlet subcoolings in a short vertical tube are derived based on the experimental data at high liquid Reynolds number. The transient CHF correlations can describe the subcooled flow boiling CHFs for the wide range of flow velocities at high liquid Reynolds number obtained in this work within ±15% difference.
Journal title :
Nuclear Engineering and Design Eslah
Serial Year :
2010
Journal title :
Nuclear Engineering and Design Eslah
Record number :
895856
Link To Document :
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