Validation of trace one-dimensional model using Phenix end of life natural circulation experiments

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

Research output: Chapter in Book/Conference proceedings/Edited volumeConference contributionScientificpeer-review

Abstract

The demonstrated technological feasibility of sodium-cooled fast reactors (SFRs) makes them stand out among the other fast reactor concepts proposed by Generation-IV International Forum (GIF) for short-term deployment. Therefore, it is necessary to develop computational tools capable of performing reliable safety analyses and plant behavior simulations under complex transient scenarios to assure SFR’s compliance with the highest safety goals. To satisfy this need, a multi-physics 3-dimensional core and system model is being developed. This will allow a more detailed representation of the physics of the plant and to anticipate more accurately plant behavior, even under wider three dimensional scenarios, such as asymmetric transients. As a first step of the tool development, a 1-dimensional thermal-hydraulic model of the French SFR Phenix was created and the Phenix end-of-life natural circulation test was simulated. The 1-dimensional model transient calculations are in good agreement, both in trend and absolute values, with the experimental data proving the value of the developed model to assess if natural convection is sufficient to cool the Phenix core after a scram. Nevertheless, some phenomena observed experimentally cannot be completely caught due to the 1-dimensional nature of the model. Thus, in view to improve the representation of plant behavior and as next step in the development of the multi-physics computational tool, the developed 1-dimensional model is being extended into a 3-dimensional thermal-hydraulics model. The coupling will be performed using the system codes TRACE-PARCS, modified to simulate more accurately sodium-cooled fast reactors.

Original languageEnglish
Title of host publication17th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2017
PublisherAssociation for Computing Machinery (ACM)
Volume2017-September
Publication statusPublished - 2017
Event17th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2017 - Xi'an, Shaanxi, China
Duration: 3 Sep 20178 Sep 2017

Conference

Conference17th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2017
CountryChina
CityXi'an, Shaanxi
Period3/09/178/09/17

Keywords

  • Model validation
  • One-dimensional model
  • PHENIX end-of-life natural circulation test
  • Sodium Fast Reactor
  • TRACE

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