Zobrazeno 1 - 10
of 22
pro vyhledávání: '"Hae Seob Choi"'
Publikováno v:
Nuclear Engineering and Technology, Vol 48, Iss 2, Pp 376-385 (2016)
A hexagonally arrayed 37-pin wire-wrapped rod bundle has been chosen to provide the experimental data of the pressure loss and flow rate in subchannels for validating subchannel analysis codes for the sodium-cooled fast reactor core thermal/hydraulic
Externí odkaz:
https://doaj.org/article/72473978cf6a41639c161ba1734caeb7
Publikováno v:
Energies, Vol 17, Iss 11, p 2714 (2024)
The experimental data of core flow distribution are indispensable for obtaining licensing and facilitating the design of fluid systems of nuclear reactors. In this study, an Advanced power reactor Core flow and Pressure (ACOP) test facility was estab
Externí odkaz:
https://doaj.org/article/a95f2c4a29614692a51094ae03036fc1
Publikováno v:
Nuclear Engineering and Technology, Vol 53, Iss 12, Pp 3892-3901 (2021)
The pressure drop of a moisture separator in a steam generator is the important design parameter to ensure the successful performance of a nuclear power plant. The moisture separators have a wide range of operating conditions based on the arrangement
Externí odkaz:
https://doaj.org/article/4b3594ca851e40bb932a015366e0d207
Publikováno v:
Nuclear Engineering and Technology, Vol 53, Iss 12, Pp 3892-3901 (2021)
The pressure drop of a moisture separator in a steam generator is the important design parameter to ensure the successful performance of a nuclear power plant. The moisture separators have a wide range of operating conditions based on the arrangement
Externí odkaz:
https://explore.openaire.eu/search/publication?articleId=doi_dedup___::da0c6e9b7391eb4b682ae08524a5d42b
https://doi.org/10.1016/j.net.2021.06.016
https://doi.org/10.1016/j.net.2021.06.016
Autor:
Hae-Seob Choi, Woo Shik Kim, Heung June Chung, Tae-Ho Lee, Joongwoon Kim, Dong-Jin Euh, Minsuk Kong
Publikováno v:
Nuclear Engineering and Design. 348:177-186
The flow characteristics at the shell side of a prototype intermediate heat exchanger (p-IHX) in a prototype generation-IV sodium-cooled fast reactor (PGSFR) were investigated experimentally in this study. The shell side of the p-IHX consisted of an
Externí odkaz:
https://explore.openaire.eu/search/publication?articleId=doi_________::864ffda5027b220ee5e270fb4edfd3b8
https://doi.org/10.1016/j.nucengdes.2019.03.013
https://doi.org/10.1016/j.nucengdes.2019.03.013
Publikováno v:
Science and Technology of Nuclear Installations, Vol 2014 (2014)
During the reflood phase of a large-break loss-of-coolant accident (LBLOCA) in a pressurized-water reactor (PWR), the fuel rods can be ballooned or rearranged owing to an increase in the temperature and internal pressure of the fuel rods. In this stu
Externí odkaz:
https://doaj.org/article/5bd0d68b060841a6908ce942e10b7c76
Autor:
Dong-Jin Euh, Sun Rock Choi, Seok-Kyu Chang, Won Sik Yang, Hae Seob Choi, Hyungmo Kim, Hyeong-Yeon Lee
Publikováno v:
Annals of Nuclear Energy. 166:108810
Flow mixing between adjacent subchannels within a wire-wrapped hexagonal fuel rod bundle affects radial heat transfer and determines the maximum cladding temperature, which is a key parameter to ensure the fuel safety margin in a sodium-cooled fast r
Externí odkaz:
https://explore.openaire.eu/search/publication?articleId=doi_________::a541d05f50af24beddc82e4eca319b53
https://doi.org/10.1016/j.anucene.2021.108810
https://doi.org/10.1016/j.anucene.2021.108810
Autor:
Yung Joo Ko, Hyeong-Yeon Lee, Hae Seob Choi, Seok-Kyu Chang, Hyungmo Kim, Sun Rock Choi, Seok Hoon Kim, Dong-Jin Euh
Publikováno v:
Annals of Nuclear Energy. 106:160-169
Hydrodynamic experiments were performed to characterize pressure drop and flow distribution in the subchannels of KAERI’s fuel bundle configuration. A 61-pin wire wrapped rod bundle was chosen as a test assembly considering the hydrodynamic similar
Externí odkaz:
https://explore.openaire.eu/search/publication?articleId=doi_________::65aebe1ed0456f7472f96eae3d0c8497
https://doi.org/10.1016/j.anucene.2017.03.024
https://doi.org/10.1016/j.anucene.2017.03.024
Autor:
Dong-Jin Euh, Jong Kuk Park, Seung Jun Lee, In-Cheol Chu, Chul-Hwa Song, Hae Seob Choi, Young Jung Youn
Publikováno v:
Nuclear Engineering and Design. 312:172-183
Experiments were performed to quantify the local bubble parameters such as void fraction, bubble velocity, interfacial area concentration, and Sauter mean diameter for the subcooled boiling flow of a refrigerant R-134a in a pressurized vertical annul
Externí odkaz:
https://explore.openaire.eu/search/publication?articleId=doi_________::32acafa003d0877183c985c8d88482a2
https://doi.org/10.1016/j.nucengdes.2016.06.027
https://doi.org/10.1016/j.nucengdes.2016.06.027
Publikováno v:
Nuclear Engineering and Design. 312:248-255
If fuel rods are ballooned or rearranged during the reflood phase of a large break loss-of-coolant accident (LBLOCA) in a pressurized-water reactor (PWR), the transient heat transfer behavior is entirely different with those of the intact fuel rods o
Externí odkaz:
https://explore.openaire.eu/search/publication?articleId=doi_________::aeca673d22085f404920820a4ae510d9
https://doi.org/10.1016/j.nucengdes.2016.08.031
https://doi.org/10.1016/j.nucengdes.2016.08.031