Abstract
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 in a HLM cooled pool reactor. The experimental campaign led to the characterization of mixed convection and thermal stratification in a HLM pool in safety relevant conditions and to the distribution of experimental data for the validation of CFD and system codes. For this purpose, several thermocouples were installed into the pool using 4 vertical supports in different circumferential position for a total of 119 thermocouples [1][2].The aim of this work is to investigate the capability of the system code RELAP5-3D© to simulate mixed convection and thermal stratification phenomena in a HLM pool in steady state conditions by comparing code results with experimental data. The pool has been simulated by a 3D component divided into 1728 volumes, 119 of which are centered in the exact position of the thermocouples. Three dimensional model of the pool is completed with a mono-dimensional nodalization of the primary main flow path. The results obtained by code simulations are compared with a steady state condition carried out in the experimental campaign. Results of axial, radial and azimuthal temperature profile into the pool are in agreement with the available experimental data Furthermore the code is able to well simulate operating conditions into the main flow path of the test section.
Highlights
TEST I simulations are carried out reducing by 5% the electrical power supplied to the fuel pin simulator (FPS), to take into account the power dissipated by joule effect in the cables and connectors, which does not contribute to the thermal power supplied to the lead-bismuth eutectic alloy (LBE) by the heat source (HS)
The experimental data are compared to the simulated values, calculated with default correlations of RELAP5-3D© v. 4.3.4 (R5-3D) and Nuclear Energy Agency (NEA) recommended correlations
The experimental campaign conducted on CIRCE-ICE facility offers additional data for the validation of R5-3D code in the frame of heavy liquid metal (HLM) pool-type reactors
Summary
Published under licence by IOP Publishing Ltd. Lead-cooled fast reactor (LFR) belongs to the six concepts selected by GIF as Generation IV systems and includes lead and lead-bismuth eutectic alloy (LBE) technologies; both coolants are chemically inert and they offer other attractive characteristics in terms of interaction with structural materials and thermodynamic features. LFR systems well respond to lesson of Fukushima accident allowing natural circulation both in nominal and accident conditions. This feature offers considerable grace time in order to cope with unprotected loss of flow transient and it permits to introduce fully passive decay heat removal system (DHR), assuring very high safety features over long periods without need for operator actions, combined with active systems [3]
Talk to us
Join us for a 30 min session where you can share your feedback and ask us any queries you have