Results 21 to 30 of about 2,651,698 (166)
Simulations of Core Damage Progression for TMI‐2 Severe Accident Using CINEMA Computer Code
Science and Technology of Nuclear Installations, Volume 2023, Issue 1, 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 capable of simulating core melt progression with thermal hydraulic analysis of the RCS (reactor coolant ...
Rae-Joon Park +6 more
wiley +1 more source
Thermal-hydraulic analysis of loss-of-cooling accident in spent fuel pool of Bushehr NPP using the RELAP5 and MELCOR [PDF]
مجله علوم و فنون هستهای, 2020 Following the Fukushima Daiichi accident, the simulation of accidents related to the Spent Fuel Pool (SFP) became more important due to the high content of long-lived radionuclides, and lack of the protection by the pressure vessel despite its low decay ...
S. Gol Narges, S.Kh. Mousavian
doaj +1 more source
Energy Science &Engineering, Volume 10, Issue 7, Page 2088-2105, July 2022., 2022 In this project, the necessity of equipping with an emergency water supply system in a supercritical circulating fluidized bed (CFB) boiler has long been a controversial topic in the industry. And a boiler electricity failure experiment was accomplished, a mathematical model established and a calculation program developed.
Yinlong Li +6 more
wiley +1 more source
The Concept of the Heat Removal System of a High‐Flux Research Reactor
Science and Technology of Nuclear Installations, Volume 2022, Issue 1, 2022., 2022 Achieving high neutron fluxes in research pressurized water reactors is directly related to the intensity of the coolant flow through the core and the pressure in it, which provides an increased saturation temperature and a margin to critical heat flux.
Vitaly Uzikov +3 more
wiley +1 more source
Energies, 2021 Magnetohydrodynamic (MHD) phenomena, due to the interaction between a magnetic field and a moving electro-conductive fluid, are crucial for the design of magnetic-confinement fusion reactors and, specifically, for the design of the breeding blanket ...
Lorenzo Melchiorri +4 more
doaj +1 more source
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 +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 +6 more
wiley +1 more source
Science and Technology of Nuclear Installations, Volume 2022, Issue 1, 2022., 2022 The use of the USNRC codes TRACE and PARCS has been considered for the coupled safety analysis of CANDU reactors. A key element of CANDU simulations is the interactions between thermal‐hydraulic and physic phenomena with the CANDU reactor regulating system (RRS). To date, no or limited development has taken place in TRACE‐PARCS in this area.
Simon Younan +2 more
wiley +1 more source
Prediction of the Loss of Feed Water Fault Signatures Using Machine Learning Techniques
Science and Technology of Nuclear Installations, Volume 2021, Issue 1, 2021., 2021 Fault diagnosis occurrence and its precise prediction in nuclear power plants are extremely important in avoiding disastrous consequences. The inherent limitations of the current fault diagnosis methods make machine learning techniques and their hybrid methodologies possible solutions to remedy this challenge.
Anselim M. Mwaura +2 more
wiley +1 more source
Research and Implementation of SVDU Simulator Based on Emulation Technology
Science and Technology of Nuclear Installations, Volume 2021, Issue 1, 2021., 2021 The safety video display unit (SVDU), as the display machine of the reactor protection system, performs the functions of displaying the reactor’s safety parameters and sending safety control commands. In order to meet the needs of nuclear power safety‐level digital control system (DCS), like designing verification, operator training, and accidental ...
Yanqun Wu +7 more
wiley +1 more source