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Apr 19 – 22, 2022
Vienna, Austria
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Coupled neutronic/thermal-hydraulic simulation of Unprotected Loss of Flow Test at Fast Flux Test Facility

Apr 20, 2022, 1:40 PM
Vienna, Austria

Vienna, Austria

ORAL Track 2. Fast Reactor Safety 2.3 Accident Analysis


Dr Mikityuk Konstantin (Paul Scherrer Institute)


Fast Flux Test Facility (FFTF) was a research Sodium-cooled Fast Reactor (SFR) operated in the 1980’s with a goal to demonstrate the inherent and passive safety characteristics of the SFR design. In the frame of this study, an attempt has been made to develop a model of the FFTF suitable to predict the system’s behavior during the Unprotected Loss of Flow (ULOF) accident. In particular, stochastic Monte Carlo code Serpent 2 is employed to build the static neutronics model of the system, results of which are further utilized to construct the neutron kinetics model of the core by employing the Purdue Advanced Reactor Core Simulator (PARCS). This model is further coupled to the plant system code TRAC/RELAP Advanced Computational Engine (TRACE) in the frame of the FAST code system in order to enable comprehensive analysis of transients within the fast reactors. Special attention is devoted to the Gas Expansion Modules (GEM) and their potential to drive the reactor core into the safe shutdown state upon occurrence of initiators of the ULOF accident. On top of that, assessment of representativeness of GEM devices in modeling the coolant boiling within the fast reactors is conducted. Furthermore, a number of sensitivity studies is performed in order to assess the performance of the various simulation tools and the modeling approaches in faithfully reproducing the experimental data. Obtained results are further compared to the measurements collected during the Loss of Flow WithOut SCRAM (LOFWOS) Test #13 performed at the FFTF. According to the outcome of the aforementioned comparison, LOFWOS Test #13 can be successfully reproduced by employing the FAST code system. Spatial reactor kinetics model is proven able to reproduce the evolution of the core parameters to a high degree of accuracy. However, by employing a fairly simple point reactor kinetics model, a conclusion is drawn that the aforementioned models perform comparatively similar in modeling the neutronics of the reactor cores of similar geometry, size and fuel composition as FFTF’s. Furthermore, it is proven that the axial power profile of the core does not suffer significant degradation in the process of activation of GEM devices. Moreover, on the basis of the comparison to the performance of the sodium plenum of the Advanced Sodium Technological Reactor for Industrial Demonstration (ASTRID), GEM devices are proven to be a good representative of the coolant boiling within the fast reactors.

Speaker's email address
Speaker's title Mr
Affiliation/Organization Paul Scherrer Institute, Switzerland
Country/Int. organization Switzerland

Primary authors

Mr Ðorđe Petrović (SCK•CEN) Dr Mikityuk Konstantin (Paul Scherrer Institute)

Presentation materials

Peer reviewing