Results 21 to 30 of about 1,365 (205)

Development of an MPS Code for Corium Behavior Analysis: 3D Alloy Melting

Science and Technology of Nuclear Installations, Volume 2022, Issue 1, 2022., 2022
The moving particle semi‐implicit (MPS) method as a Lagrangian method is attracting increasing attention in severe accident analysis. In this paper, we developed an MPS code for the corium behavior analysis with several additional models added: an improved heat transfer model to improve the calculation between different materials, an enthalpy‐based ...
Lijun Jian, Peng Yu, Jie Pei, Xiao Zeng, Yidan Yuan, Leon Cizelj   +5 more
wiley   +1 more source

Numerical Study of Natural Circulation Flow in Reactor Coolant System during a Severe Accident

Science and Technology of Nuclear Installations, Volume 2022, Issue 1, 2022., 2022
The rupturing of steam generator tubes leads to serious accidents in nuclear power plants. It causes radioactive materials to leak into the secondary system and release outside the reactor containment region. Therefore, it is important to model a technique to determine whether the natural circulation within a reactor coolant system (RCS) can cause ...
Dae Kyung Choi, Won Man Park, Sung Man Son, Kukhee Lim, Yong Jin Cho, Choengryul Choi, Iztok Tiselj   +6 more
wiley   +1 more source

Verification of the Efficacy of Passive Autocatalytic Recombiners in a Typical Pressurized Water Reactor under a Station Blackout Condition

Science and Technology of Nuclear Installations, Volume 2022, Issue 1, 2022., 2022
The presence of a stable stratified gas cloud inside the containment near or at the flammability limit may lead to deflagration or even detonation which may challenge the containment and cause a radioactive material release into the environment. To mitigate this risk, a number of approaches have been proposed, for example, containment inerting or ...
Daegwang Hong, Donghyun Cho, Jinwoo Kim, Aya Diab, Cigdem Cildag, Alexander Zulauf   +5 more
wiley   +1 more source

Analysis of loss of cooling accident in VVER-1000/V446 spent fuel pool using RELAP5 and MELCOR codes

Nuclear Engineering and Technology, 2023
Following the Fukushima nuclear disaster, the simulation of accidents in the spent fuel pool has become more noticeable. Despite the low amount of decay heat power, the consequences of the accidents in a spent fuel pool (SFP) can be severe due to the ...
Seyed Khalil Mousavian, Amir Saeed Shirani, Francesco D'Auria   +2 more
doaj   +1 more source

Study on Airborne Radionuclide Dispersion in Floating Nuclear Power Plant under the Loss‐of‐Coolant Accident

Science and Technology of Nuclear Installations, Volume 2021, Issue 1, 2021., 2021
Floating nuclear power plant is a kind of nuclear power plant on a barge moored specifically in an area of the sea. In order to study the factors influencing airborne radionuclide dispersion induced by the loss‐of‐coolant accident in floating nuclear power plant, the floating nuclear power plant platform was taken as the research object, and the ...
Shuliang Zou, Na Liu, Binhai Huang, Massimo Zucchetti   +3 more
wiley   +1 more source

Study on Release and Migration of Radionuclides under the Small Break Loss of Coolant Accident in a Marine Reactor

Science and Technology of Nuclear Installations, Volume 2021, Issue 1, 2021., 2021
Using severe accident analysis program MELCOR, the small break loss of coolant accident (SBLOCA) analysis model was established for a marine reactor. The release and migration of radionuclides were analyzed during a severe accident induced by SBLOCA. The analysis of the hydrogen source term release showed that the maximum hydrogen release amount was ...
Zhao Fang, Zou Shuliang, Liu Zejun, Xu Tao, Xu Shoulong, Huang Yan, Feng Jinjun, Enrico Zio   +7 more
wiley   +1 more source

MELCOR Analysis of a SPARC Experiment for Spray-PAR Interaction during a Hydrogen Release

Energies, 2020
A series of experiments were performed in the SPARC (spray-aerosol-recombiner-combustion) test facility to simulate a hydrogen mitigation system with the actuation of a PAR (passive auto-catalytic re-combiner) and spray system.
Hyoung Tae Kim, Jongtae Kim
doaj   +1 more source

Validation of the correlation-based aerosol model in the ISFRA sodium-cooled fast reactor safety analysis code

Nuclear Engineering and Technology, 2021
ISFRA (Integrated SFR Analysis Program for PSA) computer program has been developed for simulating the response of the PGSFR pool design with metal fuel during a severe accident.
Churl Yoon, Sung Il Kim, Sung Jin Lee, Seok Hun Kang, Chan Y. Paik   +4 more
doaj   +1 more source

Calculation and Analysis of the Source Term of the Reactor Core Based on Multivariate Analysis of Variance

Science and Technology of Nuclear Installations, Volume 2021, Issue 1, 2021., 2021
The calculation of the core source term is affected by various factors, such as fuel consumption, enrichment, specific power, and operation mode. The activity of lanthanides, fission products, and the photon source strength were calculated using the ORIGEN program. The weights of each factor were calculated by multivariate analysis of variance.
Tao Zhou, Peng Xu, Tian Qi, Xuemeng Qin, Juan Chen, Zhongguang Fu, Arkady Serikov   +6 more
wiley   +1 more source

Assessment of Accident-Tolerant Fuel with FeCrAl Cladding Behavior Using MELCOR 2.2 Based on the Results of the QUENCH-19 Experiment

Energies, 2023
To ensure the applicability of accident-tolerant fuels, their behaviors under various accidental conditions must be assessed. While the dependences of the behavior of single physical parameters can be investigated in single- or separate-effect ...
Tereza Abrman Marková, Guglielmo Lomonaco, Guido Mazzini, Martin Ševeček   +3 more
doaj   +1 more source