Zobrazeno 1 - 10
of 65
pro vyhledávání: '"Kwang-Soon Ha"'
Publikováno v:
Nuclear Engineering and Technology, Vol 54, Iss 7, Pp 2702-2713 (2022)
A steam generator tube rupture accompanying a core damage may cause the fission product to be released to environment bypassing the containment. In such an accident, the steam generator is the major path of the radioactive aerosol release.AEOLUS faci
Externí odkaz:
https://doaj.org/article/f3802095922143ccb1f6a4f2ef8dcf6e
Publikováno v:
Science and Technology of Nuclear Installations, Vol 2023 (2023)
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
Externí odkaz:
https://doaj.org/article/2c5ec480b71542f8806460f4f0d9eb27
Autor:
Han-Chul Kim, Kwang Soon Ha, Sung Joong Kim, Miro Seo, Sang-Ho Kang, Doo Yong Lee, Yong-Mann Song, Jongseong Lee, Hee-Jung Im, Chang-Sok Cho, Jei-Won Yeon, Sung Il Kim, Song-Won Cho, Jinho Song, Yong-Ho Ryu
Publikováno v:
Nuclear Engineering and Technology, Vol 49, Iss 8, Pp 1575-1588 (2017)
In order to develop a domestic research roadmap for severe accidents, a special committee was established by the Korean Nuclear Society. One of the subcommittees discussed the characteristics and the relevant technical issues in the stages of fission
Externí odkaz:
https://doaj.org/article/c4ac78dc77174eb099b3ff0a56b16a36
Publikováno v:
Nuclear Engineering and Technology, Vol 49, Iss 7, Pp 1547-1554 (2017)
VESTA (Verification of Ex-vessel corium STAbilization) and VESTA-S (-small) test facilities were constructed at the Korea Atomic Energy Research Institute in 2010 to perform various corium melt experiments. Since then, several tests have been perform
Externí odkaz:
https://doaj.org/article/7d547bd26ef44fe1831ac7e0c63a68dc
Publikováno v:
Korean Journal of Air-Conditioning and Refrigeration Engineering. 34:100-109
Externí odkaz:
https://explore.openaire.eu/search/publication?articleId=doi_________::f110d61ab2ced444a629053714a7754a
https://doi.org/10.6110/kjacr.2022.34.2.100
https://doi.org/10.6110/kjacr.2022.34.2.100
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
Externí odkaz:
https://explore.openaire.eu/search/publication?articleId=doi_dedup___::0fe5938ea5eef51fa78bde0c0badc03d
https://doi.org/10.1016/j.net.2021.06.039
https://doi.org/10.1016/j.net.2021.06.039
Publikováno v:
Nuclear Engineering and Technology, Vol 47, Iss 1, Pp 66-73 (2015)
Background: A numerical simulation was carried out to investigate the difference between internal and external heat-flux distributions at the reactor vessel wall under in-vessel retention through external reactor vessel cooling (IVR-ERVC). Methods: T
Externí odkaz:
https://doaj.org/article/f83fdf0430a84a90b23a79838decae48
Publikováno v:
Science and Technology of Nuclear Installations, Vol 2017 (2017)
A core catcher has been developed to maintain the integrity of nuclear reactor containment from molten corium during a severe accident. It uses a two-phase natural circulation for cooling molten corium. Flow in a typical core catcher is unique becaus
Externí odkaz:
https://doaj.org/article/4ffcdcc818a9438e8e96e8fe6072097c
Publikováno v:
Nuclear Engineering and Technology, Vol 46, Iss 6, Pp 797-802 (2014)
This study investigated the thermal hydraulic issues in the Containment Filtered Venting System (CFVS) for a long operating time using the MELCOR computer code. The modeling of the CFVS, including the models for pool scrubbing and the filter, was add
Externí odkaz:
https://doaj.org/article/2b679fb681c7481db111e7f84d2be776
Publikováno v:
Nuclear Engineering and Technology, Vol 45, Iss 1, Pp 21-28 (2013)
The flow and heat transfer characteristics of the ex-vessel core melt (corium) were investigated using a commercial CFD code along with the experimental data on the spreading of corium available in the literature (VULCANO VE-U7 test). In the numerica
Externí odkaz:
https://doaj.org/article/51db66d27789495a93d751565d47a4c1