Results 21 to 30 of about 5,521 (168)

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

Pool temperature stratification analysis in CIRCE-ICE facility with RELAP5-3D© model and comparison with experimental tests [PDF]

, 2017
In the frame of heavy liquid metal (HLM) technology development, CIRCE pool facility at ENEA/Brasimone Research Center was updated by installing ICE (Integral Circulation Experiments) test section which simulates the thermal behavior of a primary system ...
Caruso, G., Giannetti, F, Martelli, D, Narcisi, Vincenzo, Tarantino, M   +4 more
core   +1 more source

Development and Testing of TRACE/PARCS ECI Capability for Modelling CANDU Reactors with Reactor Regulating System Response

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, David R. Novog, Guglielmo Lomonaco   +2 more
wiley   +1 more source

Numerical analysis of temperature stratification in the CIRCE pool facility [PDF]

, 2019
In the framework of Heavy Liquid Metal (HLM) GEN IV Nuclear reactor development, the focus is in the combination of security and performance. Numerical simulations with Computational Fluid Dynamics (CFD) or system codes are useful tools to predict the ...
EDEMETTI, FRANCESCO, FERRONI, Luisa, GIANNETTI, FABIO, NARCISI, Vincenzo, TARANTINO, Mariano, TASSONE, ALESSANDRO   +5 more
core   +1 more source

SCDAP/RELAP5 independent peer review [PDF]

, 1993
The SCDAP/RELAP5 code has been developed for best-estimate transient simulation of light-water-reactor coolant systems during severe accidents. The newest version of the code is SCDAP/RELAP5/MOD3. The US Nuclear Regulatory Commission (NRC) decided that there was a need for a broad technical review of the code by recognized experts to determine overall ...
Corradini, M. L., Dhir, V. K., Haste, T. J., Heames, T. J., Jenks, R. P., Kelly, J. E., Khatib-Rahbar, M., Viskanta, R.   +7 more
openaire   +5 more sources

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, Yong-Kuo Liu, Enrico Zio   +2 more
wiley   +1 more source

Pre-test analysis of protected loss of primary pump transients in CIRCE-HERO facility [PDF]

, 2017
In the frame of LEADER project (Lead-cooled European Advanced Demonstration Reactor), a new configuration of the steam generator for ALFRED (Advanced Lead Fast Reactor European Demonstrator) was proposed. The new concept is a super-heated steam generator,
Caruso, G., Del Nevo, A, Giannetti, F, Narcisi, V, Tarantino, M   +4 more
core   +1 more source

Simulation and Analysis of Small Break LOCA for AP1000 Using RELAP5-MV and Its Comparison with NOTRUMP Code

Science and Technology of Nuclear Installations, 2017
Many reactor safety simulation codes for nuclear power plants (NPPs) have been developed. However, it is very important to evaluate these codes by testing different accident scenarios in actual plant conditions.
Eltayeb Yousif, Zhijian Zhang, Zhaofei Tian, Hao-ran Ju   +3 more
doaj   +1 more source

SCDAP/RELAP5 lower core plate model [PDF]

, 1999
The SCDAP/RELAP5 computer code is a best-estimate analysis tool for performing nuclear reactor severe accident simulations. This report describes the justification, theory, implementation, and testing of a new modeling capability which will refine the analysis of the movement of molten material from the core region to the vessel lower head.
Coryell, E. W., Griffin, F. P.
openaire   +4 more sources

Brayton cycle numerical modeling using the RELAP5-3D code, version 4.3.4

Brazilian Journal of Radiation Sciences, 2019
This work contributes to enable and develop technologies to mount fast microreactors, to generate heat and electric energy, for the purpose to warm and to supply electrically spacecraft equipment and, also, the production of nuclear space propulsion ...
Eduardo Pedroso Longhini, Paulo David de Castro Lobo, Lamartine Nogueira Frutuoso Guimarães, Francisco Antonio Braz Filho, Guilherme Borges Ribeiro   +4 more
doaj   +1 more source