Zobrazeno 1 - 10
of 56
pro vyhledávání: '"KWI SEOK HA"'
Autor:
Kwi Lim Lee, Kwi-Seok Ha, Jae-Ho Jeong, Chi-Woong Choi, Taekyeong Jeong, Sang June Ahn, Seung Won Lee, Won-Pyo Chang, Seok Hun Kang, Jaewoon Yoo
Publikováno v:
Nuclear Engineering and Technology, Vol 48, Iss 5, Pp 1071-1082 (2016)
Korea Atomic Energy Research Institute has been developing a pool-type sodium-cooled fast reactor of the Prototype Gen-IV Sodium-cooled Fast Reactor (PGSFR). To assess the effectiveness of the inherent safety features of the PGSFR, the system transie
Externí odkaz:
https://doaj.org/article/b61623e92bee486d939401d6ecb3c021
Autor:
Sang June Ahn, Kwi-Seok Ha, Won-Pyo Chang, Seok Hun Kang, Kwi Lim Lee, Chi-Woong Choi, Seung Won Lee, Jin Yoo, Jae-Ho Jeong, Taekyeong Jeong
Publikováno v:
Nuclear Engineering and Technology, Vol 48, Iss 4, Pp 952-964 (2016)
The prototype generation IV sodium-cooled fast reactor (PGSFR) has been developed by the Korea Atomic Energy Research Institute. This reactor uses sodium as a reactor coolant to transfer the core heat energy to the turbine. Sodium has chemical charac
Externí odkaz:
https://doaj.org/article/6dd18ad0b05c465fb8f636e7ba6daf05
Publikováno v:
Nuclear Engineering and Technology, Vol 47, Iss 5, Pp 523-533 (2015)
Three-dimensional flow phenomena in a wire-wrapped 37-pin fuel assembly mock-up of a Japanese loop-type sodium-cooled fast reactor, Monju, were investigated with a numerical analysis using a general-purpose commercial computational fluid dynamics cod
Externí odkaz:
https://doaj.org/article/f514e7bfa0b3494089cbe744dc2f3a43
Publikováno v:
Annals of Nuclear Energy. 131:291-296
The metal water reaction between the fuel clad and the coolant has been deemed important to safety analyses for transients for several decades. A Loss-of-coolant accident would cause the overheating of fuel clad which might involve a vigorously exoth
Externí odkaz:
https://explore.openaire.eu/search/publication?articleId=doi_________::2c6ab53fc558a8d6aaf34836e7d9ef28
https://doi.org/10.1016/j.anucene.2019.03.047
https://doi.org/10.1016/j.anucene.2019.03.047
Publikováno v:
Annals of Nuclear Energy. 115:138-153
The safety analysis of nuclear power plants has been performed using conservative approach based on conservative assumptions and boundary conditions to evaluate the safety margin of plant operation. However, this conservative approach could lead to u
Externí odkaz:
https://explore.openaire.eu/search/publication?articleId=doi_________::c227e030895b476bf579ee4f38c8b3fd
https://doi.org/10.1016/j.anucene.2018.01.033
https://doi.org/10.1016/j.anucene.2018.01.033
Publikováno v:
Nuclear Engineering and Design. 385:111499
This paper presents an innovative solution scheme that may resolve the numerical challenges of simulating the phase appearance and disappearance phenomena in multiphase flow. In this study, a least square problem for the mass and internal energy cons
Externí odkaz:
https://explore.openaire.eu/search/publication?articleId=doi_________::6ea6ed16e39b72b163dec9fb31848139
https://doi.org/10.1016/j.nucengdes.2021.111499
https://doi.org/10.1016/j.nucengdes.2021.111499
Publikováno v:
Annals of Nuclear Energy. 110:1232-1241
In this research, sensitivity and uncertainty analyses for 23 parameters were performed for unprotected loss of flow (ULOF) for the prototype Gen-IV sodium-cooled fast reactor (PGSFR) by using the parallel computing platform integrated for uncertaint
Externí odkaz:
https://explore.openaire.eu/search/publication?articleId=doi_________::615d21afec9ccf3239a2310c886e4d16
https://doi.org/10.1016/j.anucene.2017.08.041
https://doi.org/10.1016/j.anucene.2017.08.041
Publikováno v:
Nuclear Engineering and Design. 371:110935
This paper introduces a modified internal energy equation derived for multiphase flow in various flow conditions for a staggered mesh system. The pressure drop and heat dissipation terms in the internal energy conservation equation currently being de
Externí odkaz:
https://explore.openaire.eu/search/publication?articleId=doi_________::18a842d3fa746ed61487c1b20380b8d4
https://doi.org/10.1016/j.nucengdes.2020.110935
https://doi.org/10.1016/j.nucengdes.2020.110935
Publikováno v:
Nuclear Engineering and Design. 308:269-282
The MARS-LMR code has been developed by the Korea Atomic Energy Research Institute (KAERI) to analyze transients in a pool-type sodium-cooled fast reactor (SFR). Currently, KAERI is developing a prototype Gen-IV SFR (PGSFR) with metallic fuel. The de
Externí odkaz:
https://explore.openaire.eu/search/publication?articleId=doi_________::4b971bc7bbf9be2c66cc63762c7b6483
https://doi.org/10.1016/j.nucengdes.2016.08.044
https://doi.org/10.1016/j.nucengdes.2016.08.044
Publikováno v:
Annals of Nuclear Energy. 96:422-431
KAERI has been developing a prototype Gen-IV sodium cooled fast reactor (PGSFR), which is a pool type and metallic fueled fast reactor in Korea. Unprotected event is an accident accompanying by failure of a reactor protection system. PGSFR is feature
Externí odkaz:
https://explore.openaire.eu/search/publication?articleId=doi_________::0944acbc357fa27e9c47ddc12e9dfd26
https://doi.org/10.1016/j.anucene.2016.06.035
https://doi.org/10.1016/j.anucene.2016.06.035