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Dive into the research topics where L.K.H. Leung is active.

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Featured researches published by L.K.H. Leung.


Nuclear Engineering and Design | 2003

A look-up table for fully developed film-boiling heat transfer

D.C. Groeneveld; L.K.H. Leung; A.Z. Vasić; Y.J. Guo; S.C. Cheng

Abstract An improved look-up table for film-boiling heat-transfer coefficients has been derived for steam–water flow inside vertical tubes. Compared to earlier versions of the look-up table, the following improvements were made: • The database has been expanded significantly. The present database contains 77,234 film-boiling data points obtained from 36 sources. • The upper limit of the thermodynamic quality range was increased from 1.2 to 2.0. The wider range was needed as non-equilibrium effects at low flows can extend well beyond the point where the thermodynamic quality equals unity. • The surface heat flux has been replaced by the surface temperature as an independent parameter. • The new look-up table is based only on fully developed film-boiling data. • The table entries at flow conditions for which no data are available is based on the best of five different film-boiling prediction methods. The new film-boiling look-up table predicts the database for fully developed film-boiling data with an overall rms error in heat-transfer coefficient of 10.56% and an average error of 1.71%. A comparison of the prediction accuracy of the look-up table with other leading film-boiling prediction methods shows that the look-up table results in a significant improvement in prediction accuracy.


Nuclear Engineering and Design | 1989

Computation of single- and two-phase heat transfer rates suitable for water-cooled tubes and subchannels

D.C. Groeneveld; S.C. Cheng; L.K.H. Leung; C. Nguyen

Abstract A computational method for predicting heat transfer, valid for a wide range of flow conditions (from pool boiling and laminar flow conditions to highly turbulent flow), has been developed. It correctly identifies the heat transfer modes and predicts the heat transfer rates as well as transition points (such as the critical heat flux point) on the boiling curve. The computational heat transfer method consists of a combination of carefully chosen heat transfer equations for each heat transfer mode. Each of these equations has been selected because of their accuracy, wide range of application, and correct asymptotic trends. Using a mechanistically-based heat transfer logic, these equations have been combined in a convenient software package suitable for PC or mainframe application. The computational method has been thoroughly tested against many sets of experimental data. The parametric and asymptotic trends of the prediction method have been examined in detail. Correction factors are proposed for extending the use of individual predictive techniques to various geometric configurations and upstream conditions.


2014 22nd International Conference on Nuclear Engineering | 2014

Experimental Study of Heat Transfer to Supercritical Pressure Water Flowing in a 2×2 Rod Bundle

Han Wang; Qincheng Bi; Linchuan Wang; Haicai Lv; L.K.H. Leung

An experiment has recently been performed at Xi’an Jiaotong University to study the wall temperature and pressure drop at supercritical pressures with upward flow of water inside a 2×2 rod bundle. A fuel-assembly simulator with four heated rods was installed inside a square channel with rounded corner. The outer diameter of each heated rod is 8 mm with an effective heated length of 600 mm. Experimental parameters covered the pressure of 23–28 MPa, mass flux of 350–1000 kg/m2s and heat flux on the rod surface of 200–1000 kW/m2. According to the experimental data, it was found that the circumferential wall temperature distribution of a heated rod is not uniform. The temperature difference between the maximum and the minimum varies with heat flux and/or mass flux. Heat transfer characteristics of supercritical water in bundle were discussed with respect to various heat fluxes. The effect of heat flux on heat transfer in rod bundles is similar with that in tubes or annuli. In addition, flow resistance reflected in the form of pressure loss has also been studied. Experimental results showed that the total pressure drop increases with bulk enthalpy and mass flux. Four heat transfer correlations developed for supercritical pressures water were compared with the present test data. Predictions of Jackson correlation agrees closely with the experimental data.Copyright


2012 20th International Conference on Nuclear Engineering and the ASME 2012 Power Conference | 2012

Assessment of Heat-Transfer Correlations Against Experimental Data Obtained With Supercritical Water in Vertical Annuli

Zhendong Yang; Qincheng Bi; Han Wang; Gang Wu; L.K.H. Leung

Eleven correlations proposed for supercritical heat-transfer coefficients were assessed against a set of experimental data obtained recently with supercritical water flow in a vertical annular test section at Xi’an Jiaotong University. The inner heated rod of the test section had an outer diameter of 8 mm, while the outer unheated tube had an inner diameter of 16 mm (resulting in a gap size of 4 mm). The experiment covered pressure range from 23 to 28 MPa, mass-flux range from 350 to 1000 kg/m2s, and heat-flux range from 200 to 1000 kW/m2. The assessment shows relatively good agreement between predicted and experimental heat-transfer coefficients for several correlations. Some discrepancies have been observed at the region where deteriorated heat transfer, and are attributed to the modified Dittus-Boelter formulation that captures mainly the normal heat-transfer region. Overall, the Dittus-Boelter correlation is shown applicable only for the normal heat-transfer region, and significantly overpredicts the heat-transfer coefficient at the deteriorated heat-transfer region. The correlation of Bishop et al. appears valid for the current experimental database, particularly for high mass fluxes.Copyright


Nuclear Engineering and Design | 2007

The 2006 CHF look-up table

D.C. Groeneveld; Jianqiang Shan; A.Z. Vasić; L.K.H. Leung; Ahmet Durmayaz; J. Yang; S.C. Cheng; A. Tanase


Nuclear Engineering and Design | 2014

Experimental investigation of heat transfer from a 2 × 2 rod bundle to supercritical pressure water

Han Wang; Qincheng Bi; Linchuan Wang; Haicai Lv; L.K.H. Leung


Nuclear Engineering and Design | 2011

Experimental investigation of heat transfer for supercritical pressure water flowing in vertical annular channels

Wu Gang; Qincheng Bi; Zhendong Yang; Han Wang; Xiaojing Zhu; Hou Hao; L.K.H. Leung


Nuclear Engineering and Design | 2009

SCWR subchannel code ATHAS development and CANDU-SCWR analysis

Jianqiang Shan; Bo Zhang; Changying Li; L.K.H. Leung


Nuclear Engineering and Design | 2013

Issues and future direction of thermal-hydraulics research and development in nuclear power reactors

Pradip Saha; N. Aksan; J. Andersen; J. Yan; Jan-patrice Simoneau; L.K.H. Leung; F. Bertrand; Kazumi Aoto; Hideki Kamide


Progress in Nuclear Energy | 2014

Review of R&D for supercritical water cooled reactors

Thomas Schulenberg; L.K.H. Leung; Yoshiaki Oka

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D.C. Groeneveld

Atomic Energy of Canada Limited

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Han Wang

Xi'an Jiaotong University

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Qincheng Bi

Xi'an Jiaotong University

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Jianqiang Shan

Xi'an Jiaotong University

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A.Z. Vasić

Atomic Energy of Canada Limited

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Y.J. Guo

Atomic Energy of Canada Limited

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Changying Li

Xi'an Jiaotong University

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Haicai Lv

Xi'an Jiaotong University

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Linchuan Wang

Xi'an Jiaotong University

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