Zobrazeno 1 - 10
of 12
pro vyhledávání: '"Ji-Han Chun"'
Publikováno v:
Nuclear Engineering and Technology, Vol 54, Iss 2, Pp 554-564 (2022)
The natural circulation phenomena occurring in fully integrated nuclear reactors are associated with a unique formation mechanism. The phenomenon results from a structural feature of these reactors involving upward flow from the core, located in the
Externí odkaz:
https://doaj.org/article/771fc92118694aa4880e7dfe156efe9f
Publikováno v:
Nuclear Engineering and Technology, Vol 54, Iss 2, Pp 554-564 (2022)
The natural circulation phenomena occurring in fully integrated nuclear reactors are associated with a unique formation mechanism. The phenomenon results from a structural feature of these reactors involving upward flow from the core, located in the
Externí odkaz:
https://explore.openaire.eu/search/publication?articleId=doi_dedup___::1fb8688b5af625ad745f7fc4ce0343c4
https://doi.org/10.1016/j.net.2021.08.023
https://doi.org/10.1016/j.net.2021.08.023
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
Externí odkaz:
https://explore.openaire.eu/search/publication?articleId=doi_dedup___::bba78d91bab5980457bcd86298a7d50f
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
Externí odkaz:
https://explore.openaire.eu/search/publication?articleId=doi_________::658841a849b644578141270aede6a529
https://doi.org/10.1016/j.nucengdes.2014.06.030
https://doi.org/10.1016/j.nucengdes.2014.06.030
Publikováno v:
Nuclear Engineering and Design. 254:291-299
TASS/SMR-S is thermal-hydraulic design code for safety analysis of SMART plant. SBLOCA is considered as the design basis accident in the SMART. The capability of TASS/SMR-S for SBLOCA analysis was assessed using LOFT L3-7 as TMI action plan. Because
Externí odkaz:
https://explore.openaire.eu/search/publication?articleId=doi_________::29101ff971c21f40c6672942fe28069e
https://doi.org/10.1016/j.nucengdes.2012.09.029
https://doi.org/10.1016/j.nucengdes.2012.09.029
Publikováno v:
Annals of Nuclear Energy. 37:730-739
The purpose of this study is to develop a severe accident (SA) analysis method that is more reliable thorough transferring the physical status of the plant predicted by RELAP5 computer code to MAAP4 computer code. The methodology of the linkage analy
Externí odkaz:
https://explore.openaire.eu/search/publication?articleId=doi_________::305e4b2e4bcc0a3429e0c9974375197b
https://doi.org/10.1016/j.anucene.2010.01.012
https://doi.org/10.1016/j.anucene.2010.01.012
Publikováno v:
Annals of Nuclear Energy. 37:471-481
In order to identify and quantify the key parameters affecting the predictions of fission product transport and plate-out behavior in the coolant circuits of a very high temperature reactor (VHTR) system, systematic and intensive analyses were perfor
Externí odkaz:
https://explore.openaire.eu/search/publication?articleId=doi_________::d3c9431e26253f497f971672b1d23b2d
https://doi.org/10.1016/j.anucene.2010.01.009
https://doi.org/10.1016/j.anucene.2010.01.009
Autor:
Goon-Cherl Park, Ji-Han Chun
Publikováno v:
Transactions of the Korean Society of Mechanical Engineers B. 33:643-657
Externí odkaz:
https://explore.openaire.eu/search/publication?articleId=doi_________::52b2975c1aad47d842b8895d70281cde
https://doi.org/10.3795/ksme-b.2009.33.9.643
https://doi.org/10.3795/ksme-b.2009.33.9.643
Publikováno v:
Nuclear Engineering and Design. 238:3336-3347
The transient thermal–hydraulic phenomena of a DVI (Direct Vessel Injection) line break LOCA (Loss-of-Coolant Accident) in pressurized water reactor, APR1400, were investigated. In order to understand the phenomena during the LOCA transient, a redu
Externí odkaz:
https://explore.openaire.eu/search/publication?articleId=doi_________::61c8f75d0cadb2b9bd8aa1d5b15ee2ff
https://doi.org/10.1016/j.nucengdes.2008.06.018
https://doi.org/10.1016/j.nucengdes.2008.06.018
Autor:
Un-Chul Lee, Ji-Han Chun
Publikováno v:
Nuclear Engineering and Technology. 40:387-396
Many experimental analyses for annular film dryouts, which is one of the Critical Heat Flux (CHF) mechanisms, have been performed because of their importance. Numerical approaches must also be developed in order to assess the results from experiments
Externí odkaz:
https://explore.openaire.eu/search/publication?articleId=doi_________::4fc9979ac166d6520017a2342d334865
https://doi.org/10.5516/net.2008.40.5.387
https://doi.org/10.5516/net.2008.40.5.387