Results 31 to 40 of about 1,865 (177)
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
Recent Hydrodynamics Improvements to the RELAP5-3D Code [PDF]
Volume 3: Thermal Hydraulics; Current Advanced Reactors: Plant Design, Construction, Workforce and Public Acceptance, 2009 The hydrodynamics section of the RELAP5-3D computer program has been recently improved. Changes were made as follows: (1) improved turbine model, (2) spray model for the pressurizer model, (3) feedwater heater model, (4) radionuclide transport model, (5) improved pump model, and (6) compressor model. These changes will be discussed.
Riemke, Richard A. +2 more
openaire +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
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
doaj +1 more source
RELAP5/MOD3 code manual. Volume 4, Models and correlations [PDF]
, 1995 The RELAP5 code has been developed for best-estimate transient simulation of light water reactor coolant systems during postulated accidents. The code models the coupled behavior of the reactor coolant system and the core for loss-of-coolant accidents ...
core +1 more source
RELAP5 Modeling of a Siphon Break Effect on the Brazilian Multipurpose Reactor
Brazilian Journal of Radiation Sciences, 2021 This work presents the thermal-hydraulic simulation of the Brazilian Multipurpose Reactor (RMB) using the RELAP5/Mod3 code. The RMB will provide Brazil with a fundamental infrastructure for the national development on activities of the nuclear sector in ...
Humberto Vitor Soares +2 more
doaj +1 more source
Numerical Study on Laminar-Turbulent Transition Flow in Rectangular Channels of a Nuclear Reactor
Frontiers in Energy Research, 2020 Laminar-turbulent transition flow can be observed in thermal engineering applications, but the flow resistance and heat transfer characteristics are not fully understood.
Zhenying Wang +6 more
doaj +1 more source
Thermal-Hydraulics analysis of pressurized water reactor core by using single heated channel model [PDF]
Journal of Applied and Computational Mechanics, 2017 Thermal hydraulics of nuclear reactor as a basis of reactor safety has a very important role in reactor design and control. The thermal-hydraulic analysis provides input data to the reactor-physics analysis, whereas the latter gives information about the
Reza Akbari +2 more
doaj +1 more source