Zobrazeno 1 - 10
of 111
pro vyhledávání: '"Byong-Jo Yun"'
Publikováno v:
Nuclear Engineering and Technology, Vol 51, Iss 4, Pp 977-986 (2019)
In the thermal-hydraulic system codes, such as MARS and RELAP5/MOD3, the Savannah River Laboratory (SRL) model has been adopted as a subcooled boiling model. It, however, has been shown that the SRL model cannot take into account appropriately the ef
Externí odkaz:
https://doaj.org/article/11deeefbf1f34bd68bfe077e1882d352
Publikováno v:
Nuclear Engineering and Technology, Vol 47, Iss 6, Pp 669-677 (2015)
A mechanistic model for void fraction prediction with improved interfacial friction factor in nearly horizontal tubes has been proposed in connection with the development of a condensation model package for the passive auxiliary feedwater system of t
Externí odkaz:
https://doaj.org/article/d286847c5b034ae9a22e99c9fdb3e76c
Publikováno v:
Nuclear Engineering and Technology, Vol 45, Iss 6, Pp 759-766 (2013)
As passive safety features for nuclear power plants receive increasing attention, various studies have been conducted to develop safety systems for 3rd-generation (GEN-III) nuclear power plants that are driven by passive systems. The Passive Auxiliar
Externí odkaz:
https://doaj.org/article/9ce023b75acc4982bdd435f2ef93fce7
Publikováno v:
Nuclear Engineering and Technology, Vol 54, Iss 10, Pp 3962-3973 (2022)
Several critical heat flux (CHF) correlations including the look-up table in the MARS code have been assessed for the prediction of CHF in a downward-flow narrow rectangular channel. For the assessment, we built an experiment database that covers pre
Externí odkaz:
https://doaj.org/article/5f46aa058cfa4ec6b61fdb38bd8308c7
Publikováno v:
Nuclear Engineering and Technology, Vol 54, Iss 6, Pp 2297-2310 (2022)
A condensation heat transfer model is essential to accurately predict the performance of the passive containment cooling system (PCCS) during an accident in an advanced light water reactor. However, most of existing models tend to predict condensatio
Externí odkaz:
https://doaj.org/article/6a7da06073c84ba5886dd2fb84483fb5
Publikováno v:
Nuclear Engineering and Technology, Vol 54, Iss 3, Pp 1126-1135 (2022)
The MARS code has been assessed for the prediction of onset of flow instability (OFI) in a vertical channel. For assessment, we built an experiment database that consists of experiments under various geometry and thermal-hydraulic condition. It cover
Externí odkaz:
https://doaj.org/article/9ce7ccfa0d814988aa3785e5d702053b
Publikováno v:
Nuclear Engineering and Technology, Vol 53, Iss 1, Pp 322-336 (2021)
This paper presents a numerical prediction of the transient hydraulic loads acting on the tubes and external supports of a pressurized water reactor (PWR) steam generator (SG) during blowdown following a sudden feedwater line break (FWLB). A simplifi
Externí odkaz:
https://doaj.org/article/91a7f493b3f7447c90ff43f460018eef
Publikováno v:
Nuclear Engineering and Technology. 54:1126-1135
The MARS code has been assessed for the prediction of onset of flow instability (OFI) in a vertical channel. For assessment, we built an experiment database that consists of experiments under various geometry and thermal-hydraulic condition. It cover
Externí odkaz:
https://explore.openaire.eu/search/publication?articleId=doi_________::fb5ab32ae633f7352c2e5be409898f58
https://doi.org/10.1016/j.net.2021.09.013
https://doi.org/10.1016/j.net.2021.09.013
Publikováno v:
Nuclear Engineering and Technology, Vol 53, Iss 6, Pp 1796-1809 (2021)
The subchannel of a research reactor used to generate high power density is designed to be narrow and rectangular and comprises plate-type fuels operating under downward flow conditions. Critical heat flux (CHF) is a crucial parameter for estimating
Externí odkaz:
https://explore.openaire.eu/search/publication?articleId=doi_dedup___::4e5feb11be2599d3fd099c27b09a6514
https://doi.org/10.1016/j.net.2020.12.007
https://doi.org/10.1016/j.net.2020.12.007
Publikováno v:
Nuclear Engineering and Technology, Vol 53, Iss 1, Pp 322-336 (2021)
This paper presents a numerical prediction of the transient hydraulic loads acting on the tubes and external supports of a pressurized water reactor (PWR) steam generator (SG) during blowdown following a sudden feedwater line break (FWLB). A simplifi
Externí odkaz:
https://explore.openaire.eu/search/publication?articleId=doi_dedup___::bd119dc567a7ef79ade4b16bccf937fa
https://doi.org/10.1016/j.net.2020.06.020
https://doi.org/10.1016/j.net.2020.06.020