Results 31 to 40 of about 5,558 (186)
Science and Technology of Nuclear Installations, Volume 2022, Issue 1, 2022., 2022 With the increase in system complexity and operational performance requirements, nuclear energy systems are developing in the direction of intelligence and unmanned, which also requires a higher demand for its safety so that intelligent fault diagnosis and prediction have become a technology that nuclear power plants need to develop at present.
Lu Chao +5 more
wiley +1 more source
Validation of RELAP5 MOD3.3 code for Hybrid-SIT against SET and IET experimental data
Nuclear Engineering and Technology, 2020 We validated the performance of RELAP MOD3.3 code regarding the hybrid SIT with available experimental data. The concept of the hybrid SIT is to connect the pressurizer to SIT to utilize the water inside SIT in the case of SBO or SB-LOCA combined with ...
Ho Joon Yoon +3 more
doaj +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
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. +7 more
openaire +5 more sources
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. +4 more
core +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
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. +4 more
core +1 more source
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 +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
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 +5 more
core +1 more source