Zobrazeno 1 - 10
of 91
pro vyhledávání: '"L.K.H. Leung"'
Publikováno v:
Progress in Nuclear Energy. 77:282-299
The Supercritical Water-Cooled Reactor (SCWR) is a high temperature, high pressure water-cooled reactor that operates above the thermodynamic critical point (374 °C, 22.1 MPa) of water. In general terms, the conceptual designs of SCWRs can be groupe
Externí odkaz:
https://explore.openaire.eu/search/publication?articleId=doi_________::af4b678c088a642e28a7b524743b2b33
https://doi.org/10.1016/j.pnucene.2014.02.021
https://doi.org/10.1016/j.pnucene.2014.02.021
Publikováno v:
Nuclear Engineering and Design. 275:205-218
Heat transfer experiments with supercritical pressure water flowing vertically upward through a 2 × 2 rod bundle have been performed at Xi’an Jiaotong University. A fuel-assembly simulator with four heated rods was installed inside a square channe
Externí odkaz:
https://explore.openaire.eu/search/publication?articleId=doi_________::4f3510450954a2d2221230a5f92fd848
https://doi.org/10.1016/j.nucengdes.2014.04.036
https://doi.org/10.1016/j.nucengdes.2014.04.036
Issues and future direction of thermal-hydraulics research and development in nuclear power reactors
Autor:
N. Aksan, L.K.H. Leung, Jan-patrice Simoneau, F. Bertrand, Kazumi Aoto, Hideki Kamide, J. Andersen, Pradip Saha, J. Yan
Publikováno v:
Nuclear Engineering and Design. 264:3-23
The paper archives the proceedings of an expert panel discussion on the issues and future direction of thermal-hydraulic research and development in nuclear power reactors held at the NURETH-14 conference in Toronto, Canada, in September 2011. Therma
Externí odkaz:
https://explore.openaire.eu/search/publication?articleId=doi_________::e05aed2460bd7923438fa23f6c710c31
https://doi.org/10.1016/j.nucengdes.2012.07.023
https://doi.org/10.1016/j.nucengdes.2012.07.023
Autor:
E.N. Onder, L.K.H. Leung
Publikováno v:
Nuclear Engineering and Design. 264:119-125
Boiling-Length-Average (BLA) Critical Heat Flux (CHF) values for the CANFLEX 1 bundle at cross-sectional average subcooled conditions have been evaluated using the ASSERT-PV subchannel code. The predicted BLA CHF values supplement experimental BLA CH
Externí odkaz:
https://explore.openaire.eu/search/publication?articleId=doi_________::540530cbe11f407386ea87b6694fce47
https://doi.org/10.1016/j.nucengdes.2013.02.025
https://doi.org/10.1016/j.nucengdes.2013.02.025
Publikováno v:
Nuclear Engineering and Design. 241:4045-4054
An experiment has recently been completed at Xi’an Jiaotong University (XJTU) to obtain wall-temperature measurements at supercritical pressures with upward flow of water inside vertical annuli. Two annular test sections were constructed with annul
Externí odkaz:
https://explore.openaire.eu/search/publication?articleId=doi_________::2aacd8e66d6c11b4595911fe96f122de
https://doi.org/10.1016/j.nucengdes.2011.07.007
https://doi.org/10.1016/j.nucengdes.2011.07.007
Autor:
L.K.H. Leung, F.C. Dimayuga
Publikováno v:
Nuclear Engineering and Design. 240:290-298
An experiment has been performed to obtain dryout power measurements with a 37-element bundle string simulating radial power profiles of high-enriched fuel (which contains slightly higher enrichment than the CANDU fuel and is not the same as the trad
Externí odkaz:
https://explore.openaire.eu/search/publication?articleId=doi_________::e25fe8da1538838c8f3d97f098405079
https://doi.org/10.1016/j.nucengdes.2009.09.016
https://doi.org/10.1016/j.nucengdes.2009.09.016
Publikováno v:
Progress in Nuclear Energy. 51:799-804
A cooperative study has been initiated at Xi'an Jiaotong University (XJTU) with Atomic Energy of Canada Limited (AECL) to develop a subchannel code ATHAS for preliminary analyses of flow and enthalpy distributions and cladding temperatures in CANDU f
Externí odkaz:
https://explore.openaire.eu/search/publication?articleId=doi_________::4587417200013191dc8bf1ccea54d7c0
https://doi.org/10.1016/j.pnucene.2009.05.004
https://doi.org/10.1016/j.pnucene.2009.05.004
Publikováno v:
Nuclear Engineering and Design. 239:1979-1987
A subchannel code (ATHAS) is developed for preliminary analyses of flow and enthalpy distributions and cladding temperatures at supercritical water conditions. The code is applicable for transient and steady state calculations. A number of heat-trans
Externí odkaz:
https://explore.openaire.eu/search/publication?articleId=doi_________::44df1384f73672b42932ec04cfd31334
https://doi.org/10.1016/j.nucengdes.2009.05.008
https://doi.org/10.1016/j.nucengdes.2009.05.008
Autor:
L.K.H. Leung, Jianqiang Shan, Ahmet Durmayaz, A.Z. Vasic, A. Tanase, S.C. Cheng, D.C. Groeneveld, J. Yang
Publikováno v:
Nuclear Engineering and Design. 237:1909-1922
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 8 mm water-cooled tube. The 2006 CHF look-up table is based on a database containing more
Externí odkaz:
https://explore.openaire.eu/search/publication?articleId=doi_________::eda0c1723233c14a665bf4ae02a6c2bc
https://doi.org/10.1016/j.nucengdes.2007.02.014
https://doi.org/10.1016/j.nucengdes.2007.02.014
Autor:
J. Yang, L.K.H. Leung, S. W. Peng, M.A. El Nakla, A Z Vasic, D. C. Groeneveld, S.C. Cheng, Y.J. Guo
Publikováno v:
Nuclear Technology. 152:87-104
Lookup tables (LUTs) have been used widely for the prediction of critical heat flux (CHF) and film-boiling heat transfer for water-cooled tubes. LUTs are basically normalized data banks. They eliminate the need to choose between the many different CH
Externí odkaz:
https://explore.openaire.eu/search/publication?articleId=doi_________::76a07e8a8e1943b6199cdbd47a5dbfbc
https://doi.org/10.13182/nt152-87
https://doi.org/10.13182/nt152-87