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Pel, B.S.; Chen, W.H.; Chen, Y.B.; Shih, C.
Proceedings of international nuclear power plant thermal hydraulics and operations topical meeting1984
Proceedings of international nuclear power plant thermal hydraulics and operations topical meeting1984
AbstractAbstract
[en] The purpose of the study is to use extended axial nonuniform, rod-bundle critical heat flux (CHF) data points, which were released by the Heat Transfer Research Facility (HTRF) of Columbia University, to perform a statistical analysis of five publicly available CHF correlations and determine if there any way to refine these correlations. This study evaluates the performance of five CHF correlations with the COBRA IIIC/ MIT-1 thermal-hydraulic subchannel analysis code. The five correlations include the three major vendors' correlations: CE-1, BandWnumber2 and W-3. The other two correlations are more generalized and span large operating ranges, including both PWR and BWR conditions. These are Bowring's WSC-2 correlation and the EPRI-1 correlation developed at Columbia University under EPRI sponsorship). The data these correlations were evaluated against comprised a group of 1635 axial nonuniform, rod-bundle, first CHF points for PWR core geometries, spanning the pressure range of 5.14 to 17 MPa (745 to 2465 psia), the mass velocity range of 698 to 5484 kg/m2-sec (0.5145 to 4.043 Mlb/ft2-hr) and the CHF local quality range of -0.2461 to 0.6979. In the evaluations, the correlations were applied to the entire CHF points, and also applied to the points within their respective ranges. The results of this study show that all five CHF correlations can give resonable predictions when used with the COBRA IIIC/MIT-1 code based on the CHF points within their respective ranges
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Chao, J.; Chiu, C; p. D4-1-D4-10; 1984; p. D4-1-D4-10; American Nuclear Society; La Grange Park, IL (USA); International thermal hydraulics and plant operations topical meeting; Taipei, TW (China); 22-24 Oct 1984
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