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

Sara Silva Pinto Wahnon, L. Ammirabile, J. L. Kloosterman, D. Lathouwers

Research output: Contribution to journalArticleScientificpeer-review

3 Citations (Scopus)
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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.

Original languageEnglish
Pages (from-to)331-341
Number of pages11
JournalNuclear Engineering and Design
Volume331
DOIs
Publication statusPublished - 2018

Keywords

  • Hydraulics model
  • Model validation
  • Phenix end-of-life natural circulation test
  • Sodium fast reactor
  • Three-dimensional thermal
  • TRACE

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