Results 11 to 20 of about 1,536,451 (138)
Code Development in Coupled PARCS/RELAP5 for Supercritical Water Reactor
Science and Technology of Nuclear Installations, 2014 The new capability is added to the existing coupled code package PARCS/RELAP5, in order to analyze SCWR design under supercritical pressure with the separated water coolant and moderator channels. This expansion is carried out on both codes.
Po Hu, Paul Wilson
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RELAP5/MOD3.3 Code Validation with Plant Abnormal Event
Science and Technology of Nuclear Installations, 2008 Measured plant data from various abnormal events are of great importance for code validation. The purpose of the study was to validate the RELAP5/MOD3.3 Patch 03 computer code with the abnormal event which occurred at Krško Nuclear Power Plant (NPP) on ...
Andrej Prošek, Borut Mavko
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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
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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
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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
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
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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
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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
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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
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Reactivity feedback effect on loss of flow accident in PWR
Nuclear Engineering and Technology, 2018 In this work, the reactor kinetics capability is used to compute the design safety parameters in a PWR due to complete loss of coolant flow during protected and unprotected accidents.
Basma Foad +2 more
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