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
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Assessment of RELAP/SCDAPSIM/MOD3.4 Prediction Capability with Severe Fuel Damage Scoping Test
Science and Technology of Nuclear Installations, 2017 The Power Burst Facility (PBF) was designed to provide experimental data to determine the thresholds for failure during accident conditions. Thus, the PBF benchmark using severe accidental analysis codes is essential to designing reactor for current ...
Noppawan Rattanadecho +4 more
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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
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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
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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
RELAP5-3D code validation for RBMK phenomena [PDF]
, 1999 The RELAP5-3D thermal-hydraulic code was assessed against Japanese Safety Experiment Loop (SEL) and Heat Transfer Loop (HTL) tests. These tests were chosen because the phenomena present are applicable to analyses of Russian RBMK reactor designs.
Fisher, J.E., Fisher, James Ebberly
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Study of thermal-hydraulic phenomena in hot leg break LOCA of PWR systems [PDF]
, 2011 The present work cope with the execution of two post tests analysis relative to experiments executed in two different Integral test facilities. The aim of the work is to understand if the Thermal-Hydraulic system code utilized, has the capability to ...
MATTEOLI, CAMILLA
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RELAP5 Calculations of Bethsy 9.1b Test
Science and Technology of Nuclear Installations, 2012 Recently, several advanced computational tools for simulating reactor system behavior during real and hypothetical transient scenarios were developed.
Andrej Prošek
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Loss of offsite power accident analysis in a VVER-1000/V446 nuclear power plant [PDF]
Nuclear Technology and Radiation Protection, 2019 The aim of this study is to present a thermo-hydraulic analysis of the loss of offsite power accident in VVER-1000/V446 nuclear power plant using the RELAP5 code.
Esfandiari Mohsen +3 more
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RELAP5/MOD3.3 Analysis of the Loss of External Power Event with Safety Injection Actuation
Science and Technology of Nuclear Installations, 2018 The code assessment typically comprises basic tests cases, separate effects test, and integral effects tests. On the other hand, the thermal hydraulic system codes like RELAP5/MOD3.3 are primarily intended for simulation of transients and accidents in ...
Andrej Prošek, Marko Matkovič
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