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Chen Yuzhou; Yang Chungsheng; Zhao Minfu; Du Kaiwen; Zhang Shuming
Technical meeting on heat transfer, thermal-hydraulics and system design for supercritical pressure water cooled reactors. Book of abstracts2010
Technical meeting on heat transfer, thermal-hydraulics and system design for supercritical pressure water cooled reactors. Book of abstracts2010
AbstractAbstract
[en] The experiments on critical flow, heat transfer coefficient and critical heat flux have been conducted at the test loop of supercritical water in China Institute of Atomic Energy (CIAE). The major characteristics and parametric trends of these phenomena are presented, and the experimental results are compared with the calculations of existing correlations and models. For critical flow, more than 250 data points were obtained in two nozzles of 1.41 mm in diameter and 4.35 mm in length with rounded-edge and sharp-edge respectively, covering the ranges of inlet pressure of 22.1 - 29.1 MPa and inlet temperature of 38 - 474 K. The results show that in the near and beyond pseudo-critical region the thermal equilibrium is dominant, and the flow rate can be represented by the homogeneous equilibrium model reasonably. For the region of 0 < DTPC < 100 K the flow exhibits bifurcation behavior, characterized by a kind of instability, and the inlet shape of nozzle has a substantial effect on it. For the temperature well below the pseudo- critical point the flow is not at critical condition and is represented by the Bernoulli equation. For heat transfer coefficient, the experimental data were obtained in a tube of 6 mm with upward flow, covering the ranges of pressure of 10 - 23 MPa, mass flux of 288 - 1298 kg/m2s, local water temperature of 78 - 270 oC , heat flux of 0.23 - 1.18 MW/m2 and Reynolds number of 5.5 x 103 - 3.9 x 104. At both subcritical and supercritical pressures, the deterioration in heat transfer is observed at rather high Reynolds number as a result of the change in flow structure. The Dittus-Boelter type correlations e.g. Bishop's, Swenson's and Jackson's correlations, give better predictions of the heat transfer for the normal turbulent convection, but they can not predict the deteriorated heat transfer satisfactorily. For critical heat flux (CHF), the data were obtained in a tube of D = 8 mm with water flowing upward, covering the ranges of pressure of 5-18 MPa and mass flux of 0.45 - 1.5 x 103 kg/m2s. CHF increases with mass flux increasing significantly. It decreases distinctly when the pressure increases to the near critical region. The 96 CHF Look-Up Table gives reasonable prediction for higher flow, but substantial overprediction for mass flux lower than 1000 kg/m2s. (author)
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International Atomic Energy Agency, Vienna (Austria); 46 p; Jul 2010; p. 18; Technical meeting on heat transfer, thermal-hydraulics and system design for supercritical pressure water cooled reactors; Pisa (Italy); 5-8 Jul 2010; IAEA-TM--38683-13; Also available on-line: http://www-pub.iaea.org/MTCD/Meetings/PDFplus/2010/38683/38683_BookOfAbstracts.pdf
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