Some recent progress in reactor core thermal-hydraulics modelling
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[1] Xu Cheng,et al. A new correlation for post-dryout heat transfer in upward vertical flow , 2022, Nuclear Engineering and Design.
[2] Xu Cheng,et al. Studies on post-dryout heat transfer in R-134a vertical flow , 2021 .
[3] V. H. Sanchez-Espinoza,et al. Multi-scale coupling of CFD code and sub-channel code based on a generic coupling architecture , 2020, Annals of Nuclear Energy.
[4] Jinbiao Xiong,et al. High-fidelity PIV measurement of cross flow in 5 × 5 rod bundle with mixing vane grids , 2019, Nuclear Engineering and Design.
[5] Xu Cheng,et al. A new turbulent mixing modeling approach for sub-channel analysis code , 2018, Annals of Nuclear Energy.
[6] Xu Cheng,et al. Analysis of inter-channel sweeping flow in wire wrapped 19-rod bundle , 2018, Nuclear Engineering and Design.
[7] X. Cheng,et al. Mechanistic prediction of post dryout heat transfer and rewetting , 2018, Kerntechnik.
[8] Xu Cheng,et al. Analysis and modeling of post-dryout heat transfer in upward vertical flow , 2018 .
[9] Xu Cheng,et al. Sub-channel/system coupled code development and its application to SCWR-FQT loop , 2015 .
[10] B. Pang,et al. Proposal of a new type of two-phase interchannel mixing model for application to subchannel analysis of PWR conditions , 2014 .
[11] Xu Cheng,et al. Investigation on heat transfer non-uniformity in rod bundle , 2013 .
[12] A. Durmayaz,et al. The 2006 CHF look-up table , 2007 .
[13] Xu Cheng,et al. Numerical analysis of heat transfer in supercritical water cooled flow channels , 2007 .
[14] Xu Cheng,et al. CFD analysis of thermal–hydraulic behavior of heavy liquid metals in sub-channels , 2006 .
[15] X. Cheng,et al. Design analysis of core assemblies for supercritical pressure conditions , 2003 .
[16] T. H. Lee,et al. Local flow characteristics of subcooled boiling flow of water in a vertical concentric annulus , 2002 .
[17] N. Todreas,et al. Hydrodynamic models and correlations for bare and wire-wrapped hexagonal rod bundles — Bundle friction factors, subchannel friction factors and mixing parameters , 1986 .
[18] J. Weisman,et al. Prediction of critical heat flux in flow boiling at low qualities , 1983 .
[19] J. L. Wantland. ORRIBLE: a computer program for flow and temperature distribution in 19- rod LMFBR fuel subassemblies , 1974 .
[20] D. S. Rowe. COBRA IIIC: digital computer program for steady state and transient thermal-hydraulic analysis of rod bundle nuclear fuel elements , 1973 .
[21] J. C. Chen. Correlation for Boiling Heat Transfer to Saturated Fluids in Convective Flow , 1966 .
[22] S. Levy,et al. Prediction of Two-Phase Pressure Drop and Density Distribution From Mixing Length Theory , 1963 .