Zobrazeno 1 - 10
of 122
pro vyhledávání: '"D. Bestion"'
Autor:
M. Ilvonen, J. Macek, D. Lakehal, J.-M. Seynhaeve, R. Kyrki-Rajamäki, A. Martin, I. Tiselj, M. Scheuerer, F. D'Auria, D. Mazzini, B. Smith, P. Coste, D. Lucas, D. Bestion, E. Bodèle
Publikováno v:
Science and Technology of Nuclear Installations, Vol 2009 (2009)
Externí odkaz:
https://doaj.org/article/4ab492816b4842dca887ae78b8ccd881
Autor:
F. D’Auria, D. Bestion
Publikováno v:
Nuclear Technology. 208:990-1011
Akademický článek
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Autor:
M. Lanfredini, D. Bestion, F. D'Auria, N. Aydemir, S. Carnevali, P. Fillion, P. Gaillard, J.J. Jeong, M. Junk, I. Karppinen, K.D. Kim, J. Kurki, J.H. Lee, P. Schoeffel, H. Sha, T. Skorek, J.L. Vacher, G. Waddington
Publikováno v:
The 19th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-19)
The 19th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-19), Mar 2022, Brussels (online), Belgium. pp.35792
Lanfredini, M, Bestion, D, D'Auria, F, Aydemir, N, Carnevali, S, Fillion, P, Gaillard, P, Jeong, J J, Junk, M, Karppinen, I, Kim, K D, Kurki, J, Lee, J H, Schoeffel, P, Sha, H, Skorek, T, Vacher, J L & Waddington, G 2023, ' TPTF horizontal flow prediction by SYS-TH codes – Recent analyses made within the FONESYS network ', Nuclear Engineering and Design, vol. 402, 112106 . https://doi.org/10.1016/j.nucengdes.2022.112106
The 19th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-19), Mar 2022, Brussels (online), Belgium. pp.35792
Lanfredini, M, Bestion, D, D'Auria, F, Aydemir, N, Carnevali, S, Fillion, P, Gaillard, P, Jeong, J J, Junk, M, Karppinen, I, Kim, K D, Kurki, J, Lee, J H, Schoeffel, P, Sha, H, Skorek, T, Vacher, J L & Waddington, G 2023, ' TPTF horizontal flow prediction by SYS-TH codes – Recent analyses made within the FONESYS network ', Nuclear Engineering and Design, vol. 402, 112106 . https://doi.org/10.1016/j.nucengdes.2022.112106
The horizontal stratification occurrence in water cooled nuclear reactors (WCNR) plays an important role in many transients such as loss of coolant accidents and loss of residual heat removal by influencing the liquid mass repartition, the natural ci
Externí odkaz:
https://explore.openaire.eu/search/publication?articleId=doi_dedup___::d98c4b5a8349883db4a6a253b6ef4256
https://cea.hal.science/cea-04098083
https://cea.hal.science/cea-04098083
Akademický článek
Tento výsledek nelze pro nepřihlášené uživatele zobrazit.
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Autor:
P. Gaillard, N. Aksan, D. Bestion, M. Lanfredini, Joona Kurki, Jaeseok Heo, Philippe Fillion, A. Shen, Ismo Karppinen, D. Wang, J. L. Vacher, F. D'Auria, Kwang-Rag Kim, L. Liu
Publikováno v:
Lanfredini, M, Bestion, D, D'Auria, F, Aksan, N, Fillion, P, Gaillard, P, Heo, J, Karppinen, I, Kim, K D, Kurki, J, Liu, L, Shen, A, Vacher, J L & Wang, D 2020, ' Critical flow prediction by system codes – Recent analyses made within the FONESYS network ', Nuclear Engineering and Design, vol. 366, 110731 . https://doi.org/10.1016/j.nucengdes.2020.110731
A benchmark activity on Two-Phase Critical Flow (TPCF) prediction was conducted in the framework of the Forum & Network of System Thermal-Hydraulics Nuclear Reactor Thermal-Hydraulics (FONESYS). FONESYS is a network among code developers who share th
Externí odkaz:
https://explore.openaire.eu/search/publication?articleId=doi_dedup___::98c4782aad03a36190c36a1d358f1e5f
http://hdl.handle.net/11568/1064167
http://hdl.handle.net/11568/1064167
Autor:
D. Bestion
Publikováno v:
Nuclear Engineering and Design. 312:12-29
System thermalhydraulic investigations of Design Basis Accident require several tools and methods including the Process Identification and Ranking Table, the scaling, experiment analysis, modelling, code development, code Validation and Verification,
Publikováno v:
Annals of Nuclear Energy. 83:283-297
The SENSAS – Stabilisation de l’Ebullition sodium(N) en Sortie d’ASsemblage – test bench has been settled in the frame of an R&D program on GEN IV sodium Fast Reactor (SFR) aiming to assess the phenomenology in case of Na boiling during a pos
Autor:
D. Bestion
Publikováno v:
Nuclear Engineering and Design. 279:116-125
System thermalhydraulic codes model all two-phase flow regimes but they are limited to a macroscopic description. Two-phase CFD tools predict two-phase flow with a much finer space resolution but the current modelling capabilities are limited to disp
Autor:
N. Aksan, D. Bestion, F. D’Auria, G.M. Galassi, H. Glaeser, Y. Hassan, J.J. Jeong, P.L. Kirillov, C. Morel, H. Ninokata, F. Reventos, U.S. Rohatgi, R.R. Schultz, K. Umminger
Externí odkaz:
https://explore.openaire.eu/search/publication?articleId=doi_________::46204a157af86a35e0973cc1dd538fb4
https://doi.org/10.1016/b978-0-08-100662-7.09987-5
https://doi.org/10.1016/b978-0-08-100662-7.09987-5