Results 1 to 10 of about 12,514 (165)

Evaluation of heat-flux distribution at the inner and outer reactor vessel walls under the in-vessel retention through external reactor vessel cooling condition

Nuclear Engineering and Technology, 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: Total
Jaehoon Jung, Sang Mo An, Kwang Soon Ha, Hwan Yeol Kim   +3 more
doaj   +3 more sources

Reactor Vessel Water Level Estimation During Severe Accidents Using Cascaded Fuzzy Neural Networks

Nuclear Engineering and Technology, 2016
Global concern and interest in the safety of nuclear power plants have increased considerably since the Fukushima accident. In the event of a severe accident, the reactor vessel water level cannot be measured.
Dong Yeong Kim, Kwae Hwan Yoo, Geon Pil Choi, Ju Hyun Back, Man Gyun Na   +4 more
doaj   +3 more sources

Study on the influences of RPV deformation on CHF under IVR conditions

Frontiers in Energy Research, 2023
After the reactor core melt, the thermal impact of high temperature molten mixture to the lower head of reactor pressure vessel (RPV) will cause the shape change of cooling channel constituted by reactor pressure vessel outer surface and other supporting
Shuang Wen, Shuang Wen, Xinli Gao, Xinli Gao, Rui Zhang, Kebin Fan, Bin Jia, Bin Jia, Jianping Jing, Fudong Liu, Fudong Liu, Xinlu Tian, Xinlu Tian, Xinlu Tian   +13 more
doaj   +1 more source

Test of Critical Heat Flux on Outer Surface of CAP1400 Pressurized Reactor Vessel

Shanghai Jiaotong Daxue xuebao, 2022
A full height sliced test facility is established based on the modeling process. The heating test section is manufactured by adopting the explosive welding technology, and the surface of carbon steel can simulate the actual surface characteristic of a ...
SHI Guobao, ZHENG Mingguang, ZHANG Kun, KUANG Bo, LIU Pengfei
doaj   +1 more source

Investigation of seismic responses of reactor vessel and internals for beyond-design basis earthquake using elasto-plastic time history analysis

Nuclear Engineering and Technology, 2021
Existing elastic analysis methods cannot be adhered to in order to assess the structural integrity of a reactor vessel and internals for a beyond design basis earthquake.
Sang-Jeong Lee, Eun-ho Lee, Changkyun Lee, No-Cheol Park, Youngin Choi, Changsik Oh   +5 more
doaj   +1 more source

Transient heat transfer and crust evolution during debris bed melting process in the hypothetical severe accident of HPR1000

Nuclear Engineering and Technology, 2023
In the late in-vessel phase of a nuclear reactor severe accident, the internal heat transfer and crust evolution during the debris bed melting process have important effects on the thermal load distribution along the vessel wall, and further affect the ...
Chao Lv, Gen Li, Jinchen Gao, Jinshi Wang, Junjie Yan   +4 more
doaj   +1 more source

PREDICTION OF THE REACTOR VESSEL WATER LEVEL USING FUZZY NEURAL NETWORKS IN SEVERE ACCIDENT CIRCUMSTANCES OF NPPS

Nuclear Engineering and Technology, 2014
Safety-related parameters are very important for confirming the status of a nuclear power plant. In particular, the reactor vessel water level has a direct impact on the safety fortress by confirming reactor core cooling.
SOON HO PARK, DAE SEOP KIM, JAE HWAN KIM, MAN GYUN NA   +3 more
doaj   +1 more source

Vessel failure sensitivities of an advanced reactor for SBLOCA

Nuclear Engineering and Technology, 2020
Plant-specific analyses of an advanced reactor have been performed to assure the structural integrity of the reactor pressure vessel during transient conditions, which are expected to initiate pressurized thermal shock (PTS) events.
Myung Jo Jhung, Chang-Sik Oh, Youngin Choi, Sung-Sik Kang   +3 more
doaj   +1 more source

Numerical study on thermal-hydraulics of external reactor vessel cooling in high-power reactor using MARS-KS1.5 code: CFD-aided estimation of natural circulation flow rate

Nuclear Engineering and Technology, 2022
This paper presents a numerical investigation of two-phase natural circulation flows established when external reactor vessel cooling is applied to a severe accident of the APR1400 reactor for the in-vessel retention of the core melt.
Min Seop Song, Il Woong Park, Eung Soo Kim, Yeon-Gun Lee   +3 more
doaj   +1 more source

Analysis on the discharge characteristics and spreading behavior of an ex-vessel core melt in the SMART

Nuclear Engineering and Technology, 2022
The aim of this research is to analyze the characteristics of a core melt discharged from the reactor vessel and the spreading behavior the core melt in the reactor cavity of the SMART.
Sang Ho Kim, Jaehyun Ham, Byeonghee Lee, Sung Il Kim, Hwan Yeol Kim, Rae-Joon Park, Jaehoon Jung   +6 more
doaj   +1 more source