Zobrazeno 1 - 10
of 44
pro vyhledávání: '"Bub Dong Chung"'
Moving reactor model for the MULTID components of the system thermal-hydraulic analysis code MARS-KS
Publikováno v:
Nuclear Engineering and Technology, Vol 54, Iss 11, Pp 4373-4391 (2022)
Marine reactor systems experience platform movement, and therefore, the system thermal-hydraulic analysis code needs to reflect the motion effect on the fluid to evaluate reactor safety. A moving reactor model for MARS-KS was developed to simulate th
Externí odkaz:
https://doaj.org/article/025a4d2f66a949c8b2a577c243a2626c
Publikováno v:
Nuclear Engineering and Technology, Vol 46, Iss 4, Pp 481-488 (2014)
An analysis of the uncertainty quantification related to LBLOCA using the Monte-Carlo calculation has been performed and compared with the tolerance level determined by the Wilks’ formula. The uncertainty range and distribution of each input parame
Externí odkaz:
https://doaj.org/article/6b07ef5acef04e40b07a933e53e71023
Publikováno v:
Science and Technology of Nuclear Installations.
As an integrated computer code development for severe accident sequence analysis in Korea, CINEMA has been developing from an initiation event to a containment failure. The CINEMA computer code is composed of CSPACE, SACAP, and SIRIUS, which are capa
Autor:
YuJung Choi, Sung Won Bae, Dong-Gun Son, Bub-Dong Chung, JinHo Song, JunHo Bae, Kwang-Soon Ha
Publikováno v:
Nuclear Engineering and Technology, Vol 53, Iss 12, Pp 3990-4002 (2021)
CSPACE (Core meltdown, Safety and Performance Analysis CodE for nuclear power plants) for a simulation of severe accident progression in a Pressurized Water Reactor (PWR) is developed by coupling of verified system thermal hydraulic code of SPACE (Sa
Publikováno v:
Transactions of the Korean Society of Mechanical Engineers - B. 43:531-540
Autor:
Agnès de Crécy, Jinzhao Zhang, Damar Canggih Wicaksono, Victor Sanchez, J. Freixa, T. Skorek, Francesc Reventos, Yuri Shvestov, Deog Yeon Oh, Alexander Falkov, Rostislav Pernica, A. Kovtonyuk, Milos Kyncl, Torsti Alku, Fabrice Fouet, Alexey Gusev, Dong Li, Rafael Mendizábal, Christine Sarrette, Elsa de Alfonso, Omar Zerkak, Pierre Probst, Joona Kurki, Andreas Pautz, Wadim Jäger, Bub Dong Chung, Tran Tranh Tram, Jean Baccou, Xiaojing Liu, Alessandro Petruzzi
Publikováno v:
Recercat. Dipósit de la Recerca de Catalunya
instname
UPCommons. Portal del coneixement obert de la UPC
Universitat Politècnica de Catalunya (UPC)
Skorek, T, de Crécy, A, Kovtonyuk, A, Petruzzi, A, Mendizábal, R, de Alfonso, E, Reventós, F, Freixa, J, Sarrette, C, Kyncl, M, Pernica, R, Baccou, J, Fouet, F, Probst, P, Chung, B D, Tram, T T, Oh, D Y, Gusev, A, Falkov, A, Shvestov, Y, Li, D, Liu, X, Zhang, J, Alku, T, Kurki, J, Jäger, W, Sánchez, V, Wicaksono, D, Zerkak, O & Pautz, A 2019, ' Quantification of the uncertainty of the physical models in the system thermal-hydraulic codes : PREMIUM benchmark ', Nuclear Engineering and Design, vol. 354, 110199 . https://doi.org/10.1016/j.nucengdes.2019.110199
Nuclear Engineering and Design
Nuclear Engineering and Design, Elsevier, 2019, 354, pp.110199. ⟨10.1016/j.nucengdes.2019.110199⟩
Nuclear Engineering and Design, 2019, 354, pp.110199. ⟨10.1016/j.nucengdes.2019.110199⟩
instname
UPCommons. Portal del coneixement obert de la UPC
Universitat Politècnica de Catalunya (UPC)
Skorek, T, de Crécy, A, Kovtonyuk, A, Petruzzi, A, Mendizábal, R, de Alfonso, E, Reventós, F, Freixa, J, Sarrette, C, Kyncl, M, Pernica, R, Baccou, J, Fouet, F, Probst, P, Chung, B D, Tram, T T, Oh, D Y, Gusev, A, Falkov, A, Shvestov, Y, Li, D, Liu, X, Zhang, J, Alku, T, Kurki, J, Jäger, W, Sánchez, V, Wicaksono, D, Zerkak, O & Pautz, A 2019, ' Quantification of the uncertainty of the physical models in the system thermal-hydraulic codes : PREMIUM benchmark ', Nuclear Engineering and Design, vol. 354, 110199 . https://doi.org/10.1016/j.nucengdes.2019.110199
Nuclear Engineering and Design
Nuclear Engineering and Design, Elsevier, 2019, 354, pp.110199. ⟨10.1016/j.nucengdes.2019.110199⟩
Nuclear Engineering and Design, 2019, 354, pp.110199. ⟨10.1016/j.nucengdes.2019.110199⟩
International audience; PREMIUM (Post BEMUSE Reflood Models Input Uncertainty Methods) was an activity launched with the aim ofpushing forward the methods of quantification of physical model uncertainties in thermal-hydraulic codes. Thebenchmark PREM
Externí odkaz:
https://explore.openaire.eu/search/publication?articleId=doi_dedup___::034e92523802183d7d777b90e0a0b627
https://hdl.handle.net/2117/173446
https://hdl.handle.net/2117/173446
Publikováno v:
Nuclear Engineering and Design. 289:287-295
The FCM fueled cores proposed as an accident tolerant concept is assessed against the design-basis-accident (DBA) and the beyond-DBA (BDBA) scenarios using MARS code. A thermal conductivity model of FCM fuel is incorporated in the MARS code to take i
Autor:
S. Lutsanych, E. Coscarelli, Bub Dong Chung, Francesco Saverio D'Auria, I. Horvatovic, A. Kovtonyuk, P. Emonot, H. Austregesilo, Kwang-Rag Kim, Seung-Hoon Ahn, Alessandro Petruzzi, J.L. Gandrille, Markku Hänninen, J.Y. Sauvage, N. Aksan, D. Bestion
Publikováno v:
Ahn, S H, Aksan, N, Austregesilo, H, Bestion, D, Chung, B D, Coscarelli, E, D'Auria, F, Emonot, P, Gandrille, J L, Sauvage, J Y, Hänninen, M, Horvatovic, I, Kim, K D, Kovtonyuk, A, Lutsanych, S & Petruzzi, A 2017, ' Hyperbolicity and numerics in SYS-TH codes : The FONESYS point of view ', Nuclear Engineering and Design, vol. 322, pp. 227-239 . https://doi.org/10.1016/j.nucengdes.2017.06.033
The paper provides a brief overview on the state of the art of the numerical modeling in present system thermal-hydraulic codes in the framework of the Forum & Network of System Thermal-Hydraulics Codes in Nuclear Reactor Thermal-Hydraulics (FONESYS)
Externí odkaz:
https://explore.openaire.eu/search/publication?articleId=doi_dedup___::50584eb7168e446135d753728e53fadc
http://hdl.handle.net/11568/887262
http://hdl.handle.net/11568/887262
Autor:
Ji-Han Chun, Kyoo-Hwan Bae, Bub-Dong Chung, Keung-Koo Kim, Guy-Hyung Lee, Young-Jong Chung, Young In Kim
Publikováno v:
Nuclear Engineering and Design. 277:138-145
A safety evaluation of SMART adopting a fully passive safety system was performed for a small-break loss of coolant accident (SBLOCA) with various break locations and sizes using the MARS code. It is necessary for SMART, adopting a fully passive safe
Publikováno v:
Nuclear Engineering and Technology, Vol 46, Iss 4, Pp 481-488 (2014)
An analysis of the uncertainty quantification related to LBLOCA using the Monte-Carlo calculation has been performed and compared with the tolerance level determined by the Wilks’ formula. The uncertainty range and distribution of each input parame