Title: Multi-physics models for design basis accident analysis of Sodium Fast Reactors. Part I: Validation of three-dimensional TRACE thermal-hydraulics model using Phenix end-of-life experiments
Authors: SILVA PINTO WAHNON SARAAMMIRABILE LUCAKLOOSTERMANN JAN-LEENLATHOUWERS D.
Citation: NUCLEAR ENGINEERING AND DESIGN vol. 331 p. 331-341
Publisher: ELSEVIER SCIENCE SA
Publication Year: 2018
JRC N°: JRC107618
ISSN: 0029-5493
URI: http://publications.jrc.ec.europa.eu/repository/handle/JRC107618
DOI: 10.1016/j.nucengdes.2018.02.038
Type: Articles in periodicals and books
Abstract: The demonstrated technological feasibility of Sodium-cooled Fast Reactors (SFRs) makes them stand out among the other reactor concepts proposed by Generation IV International Forum (GIF) for short-term deployment. The availability of reliable computational tools in support of safety analyses and plant simulations under complex transient scenarios is essential to assure SFR’s compliance with the highest safety goals. Answering this need, a multi-physics three-dimensional core and system model is being developed to enable a more detailed representation of the physics of the plant and to anticipate more accurately plant behaviour, even under wider three-dimensional scenarios, such as asymmetric transients. The coupling will be performed using the U.S.NRC system codes TRACE-PARCS, modified to simulate more accurately when using sodium as coolant. The publicly available end-of-life tests conducted in the French SFR Phenix were chosen as baseline to perform a first validation of the computational model. The development of the tool started with a three-dimensional thermal-hydraulic nodal system of Phenix using the TRACE system code. The system simulates the Phenix end-of-life natural circulation test and the result have been compared with published experimental and benchmark results. The main physical phenomena of the 3 phases of the transient (rise in temperature in the low part of the reactor vessel, establishment of natural convection and subsequent cooling of the lower and upper part of the vessel) are predicted by the developed nodal system. More specifically, the analysis of parameters such as Intermediate Heat Exchangers (IHX), primary pumps and core temperatures, shows that the developed system is able to predict and study natural convection phenomena in Phenix-type reactors. The three-dimensional nodal system is able to clearly illustrate the existing thermal stratification in the hot pool, which is neglected by one-dimensional systems and enables the modelling of thermal hydraulic asymmetric behaviour, as it is shown by the uneven flow distribution in Phenix’s primary IHXs as they are asymmetrically located in the reactor vessel.
JRC Directorate:Nuclear Safety and Security

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