Abstract
The present work deals with simulations carried out at the University of Pisa by using the System Thermal Hydraulics code RELAP5/Mod3.3 to support the experimental campaign conducted at the ENEA (Energia Nucleare ed Energie Alternative) Brasimone Research Centre on the CIRColazione Eutettico—Heavy liquid mEtal pRessurized water cOoled tubes (CIRCE-HERO) facility. CIRCE is an integral effect pool type facility dedicated to the study of innovative nuclear systems and cooled by heavy liquid metal, while HERO is a heat exchanger heavy liquid metal/ pressurized cooling water system hosted inside the CIRCE facility. Beside the H2020 project Multi-Purpose Hybrid Research Reactor for High-Tech Applications (MYRRHA) Research and Transmutation Endeavour (MYRTE), a series of experiments were performed with the CIRCE-HERO facility, for both nominal steady-state settings and accidental scenarios. In this framework, the RELAP5/Mod3.3 code was used to simulate the experimental tests assessing the heat losses of the facility and analyzing the thermal hydraulics phenomena occurring during the postulated Protected Loss Of Flow Accident (PLOFA). The modified version Mod. 3.3 of the source code RELAP5 was developed by the University of Pisa to include the updated thermo–physical properties and convective heat transfer correlations suitable for heavy liquid metals. After reproducing the facility through an accurate nodalization, boundary conditions were applied according to the experiments. Then, the PLOFA scenarios were reproduced by implementing the information obtained by the experimental campaign. Sensitivity analyses of the main parameters affecting the thermofluidynamics of the Lead-Bismuth Eutectic (LBE) were carried out. In the simulated scenario, the LBE mass flow rate strongly depends on the injected argon flow time trend. The numerical results are in agreement with the experimental data, however further investigations are planned to analyze the complex phenomena involved.
Highlights
The Lead-Cooled Fast Reactor (LFR) is one of the six nuclear reactor concepts selected by the Generation IV International Forum (GIF) to sustain the R&D of the generation of nuclear energy systems [1]
The LFRs are cooled by Heavy Liquid Metals (HLM) such as lead or Lead-Bismuth Eutectic (LBE), employing a fast-neutron spectrum and a closed fuel cycle for efficient conversion of fertile uranium and management of actinides
The version of RELAP5/Mod3.3 employed for this purpose was modified at Laboratory of Numerical Simulation for Nuclear Thermohydraulics of the University of Pisa by including the thermodynamic properties of lead and LBE, adding the updated transport properties for viscosity, thermal conductivity, surface tension and specific convective heat transfer correlations for liquid metals [9]
Summary
The Lead-Cooled Fast Reactor (LFR) is one of the six nuclear reactor concepts selected by the Generation IV International Forum (GIF) to sustain the R&D of the generation of nuclear energy systems [1]. Reactor for High-Tech Applications (MYRRHA) research facility [3] Inside these two projects, both experimental campaigns and numerical analyses were planned. In the present work the STH code RELAP5/Mod3.3 [5,6] was used to support the experimental campaign conducted at the ENEA (Energia Nucleare ed Energie Alternative) Brasimone Research Centre on the CIRColazione Eutettico- Heavy liquid mEtal pRessurized water cOoled tubes (CIRCE-HERO). The version of RELAP5/Mod3.3 employed for this purpose was modified at Laboratory of Numerical Simulation for Nuclear Thermohydraulics of the University of Pisa by including the thermodynamic properties of lead and LBE, adding the updated transport properties for viscosity, thermal conductivity, surface tension and specific convective heat transfer correlations for liquid metals [9]. Afterwards, sensitivity analyses of the main parameters affecting the thermo–fluid-dynamics of the Lead-Bismuth Eutectic (LBE) were carried out
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