Search Results - "Leung, L.K.H."

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  1. 1

    Assessment of CHF characteristics at subcooled conditions for the CANFLEX bundle by Onder, E.N., Leung, L.K.H.

    Published in Nuclear engineering and design (01-11-2013)
    “…Boiling-Length-Average (BLA) Critical Heat Flux (CHF) values for the CANFLEX11CANFLEX® (CANdu FLEXible) is a registered trademark of Atomic Energy of Canada…”
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    Journal Article
  2. 2

    The 1995 look-up table for critical heat flux in tubes by Groeneveld, D.C., Leung, L.K.H., Kirillov, P.L., Bobkov, V.P., Smogalev, I.P., Vinogradov, V.N., Huang, X.C., Royer, E.

    Published in Nuclear engineering and design (01-06-1996)
    “…An updated version of the look-up table for critical heat flux (CHF) has been developed jointly by AECL Research (Canada) and IPPE (Obninsk, Russia). It is…”
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    Journal Article
  3. 3

    A mechanistic bubble crowding model for predicting critical heat flux in subchannels of a bundle by Liu, Yang, Liu, Wei, Shan, Jianqiang, Zhang, Bo, Leung, L.K.H.

    Published in Annals of nuclear energy (01-03-2020)
    “…•A new mechanistic model has been developed for predicting CHF of a bundle.•Detailed descriptions of various models of bundle effects on CHF.•The application…”
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    Journal Article
  4. 4

    The 2006 CHF look-up table by Groeneveld, D.C., Shan, J.Q., Vasić, A.Z., Leung, L.K.H., Durmayaz, A., Yang, J., Cheng, S.C., Tanase, A.

    Published in Nuclear engineering and design (01-09-2007)
    “…CHF look-up tables are used widely for the prediction of the critical heat flux (CHF). The CHF look-up table is basically a normalized data bank for a vertical…”
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    Journal Article Conference Proceeding
  5. 5

    A phenomenological CHF model for mixing-vane spacers in a subchannel of a rod bundle by Liu, Yang, Dong, Siying, Shan, Jianqiang, Zhang, Bo, Jiang, Li, Liu, Wei, Leung, L.K.H.

    Published in Annals of nuclear energy (01-07-2020)
    “…•A phenomenological CHF model for mixing-vane spacers is proposed.•The new model can accurately reflect the influence of spacers on CHF prediction.•Based on…”
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    Journal Article
  6. 6

    Experimental investigation of heat transfer for supercritical pressure water flowing in vertical annular channels by Gang, Wu, Bi, Qincheng, Yang, Zhendong, Wang, Han, Zhu, Xiaojing, Hao, Hou, Leung, L.K.H.

    Published in Nuclear engineering and design (01-09-2011)
    “…► Two annular test sections were constructed with annular gaps of 4 and 6 mm. ► Two heat transfer regions have been observed: normal and deteriorated heat…”
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    Journal Article
  7. 7

    Overview of methods to increase dryout power in CANDU fuel bundles by Groeneveld, D.C., Leung, L.K.H., Park, J.H.

    Published in Nuclear engineering and design (01-06-2015)
    “…•Small changes in bundle geometry can have noticeable effects on the bundle CHF.•Rod spacing devices can results in increases of over 200% in CHF.•CHF…”
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    Journal Article
  8. 8

    Measurements of critical heat flux in CANDU 37-element bundle with a steep variation in radial power profile by Leung, L.K.H., Dimayuga, F.C.

    Published in Nuclear engineering and design (01-02-2010)
    “…An experiment has been performed to obtain dryout power measurements with a 37-element bundle string simulating radial power profiles of high-enriched fuel…”
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    Journal Article Conference Proceeding
  9. 9

    A look-up table for fully developed film-boiling heat transfer by Groeneveld, D.C., Leung, L.K.H., Vasic’, A.Z., Guo, Y.J., Cheng, S.C.

    Published in Nuclear engineering and design (01-10-2003)
    “…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…”
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    Journal Article
  10. 10

    An experimental and analytical study of the effect of axial power profile on CHF by Yang, J., Groeneveld, D.C., Leung, L.K.H., Cheng, S.C., Nakla, M.A. El

    Published in Nuclear engineering and design (01-07-2006)
    “…The effect of axial heat flux distribution (AFD) on the critical heat flux (CHF) was investigated. CHF measurements were obtained with HFC-134a cooled vertical…”
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    Journal Article
  11. 11

    Pressure drops for steam and water flow in heated tubes by Leung, L.K.H., Groeneveld, D.C., Teyssedou, A., Aubé, F.

    Published in Nuclear engineering and design (01-01-2005)
    “…Measurements of pressure drop for a steam and water flow inside several heated tubes were obtained at two different laboratories (Chalk River Laboratories and…”
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    Journal Article
  12. 12

    Thermalhydraulics studies examining the feasibility for introducing slightly enriched uranium fuel into the Embalse CANDU reactor by Leung, L.K.H., Serrano, P., Schivo, M., Parrondo, A., Guo, Y., Mazzantini, O., Oh, D., Higa, M., Khatchikian, F., Mollerach, R., Fink, J.

    Published in Nuclear engineering and design (01-09-2007)
    “…A joint study on the technical feasibility of using 0.9% slightly enriched uranium (SEU) fuel in the Embalse CANDU reactor was performed by Atomic Energy of…”
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    Journal Article Conference Proceeding
  13. 13

    Prediction of the obstacle effect on film-boiling heat transfer by Leung, L.K.H., Groeneveld, D.C., Zhang, J.

    Published in Nuclear engineering and design (01-03-2005)
    “…A correlation has been developed to account for the effect of obstacles (simulating the spacing devices in bundles) on heat transfer in dispersed-flow film…”
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    Journal Article
  14. 14

    Comparison of CHF measurements in horizontal and vertical tubes cooled with R-134a by Pioro, I.L, Groeneveld, D.C, Leung, L.K.H, Doerffer, S.S, Cheng, S.C, Antoshko, Yu.V, Guo, Y, Vasić, A

    “…An experimental study of the critical heat flux (CHF) in horizontal and vertical tubes cooled with R-134a has been completed. The investigated ranges of flow…”
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    Journal Article