Since the Lead-cooled Fast Reactor (LFR) has been conceptualized in the frame of GEN IV International Forum (GIF), ENEA is strongly involved in the HLM technology development. In particular, several experimental campaign employing HLM loop and pool facilities (CIRCE [1], NACIE [2], HELENA [3], HERO [4]) are carried out in order to support HLM technologies development. In this frame, suitable experiments were carried out on the CIRCE pool facility refurbished with the Integral Circulation Experiment (ICE) test section in order to investigate the thermal hydraulics and the heat transfer in grid spaced Fuel Pin Bundle cooled by liquid metal providing, among the others aim, experimental data in support of codes validation for the European fast reactor development. The study of thermal stratification in large pool reactor is relevant in the design of HLM nuclear reactor especially for safety issue. Thermal stratification should induce thermo-mechanical stresses on the structures and in accidental scenario conditions, could opposes to the establishment of natural circulation which is a fundamental aspect for the achievements of safety goals required in the GEN-IV roadmap. In the present work, a Protected Loss of Heat Sink with Loss Of Flow (PLOHS+LOF) scenario is experimentally simulated and the mixed convection with thermal stratification phenomena is investigated during the simulated transient, as foreseen in the frame of Horizon 2020 SESAME project [5]. A full characterization of thermal stratification inside the pool is presented, and the main results gained during the run are reported. The two tests named A (20 h) and B (6 h) here reported, essentially differs for the power supplied to the fuel bundle during the full power run (800 kW and 600 kW respectively). After the transition to natural circulation conditions, the power supplied to the bundle is decreased to about 30 kW simulating the decay heat. Finally the Nusselt number for the central subchannel of the fuel bundle simulator (FPS) is evaluated and compared with values obtained from Ushakov and Mikityuk correlations [6-7]. Copyright © 2016 by ASME.

Experimental activity for the investigation of mixing and thermal stratification phenomena in the circe pool facility

Di Piazza, I.;Tarantino, M.
2016

Abstract

Since the Lead-cooled Fast Reactor (LFR) has been conceptualized in the frame of GEN IV International Forum (GIF), ENEA is strongly involved in the HLM technology development. In particular, several experimental campaign employing HLM loop and pool facilities (CIRCE [1], NACIE [2], HELENA [3], HERO [4]) are carried out in order to support HLM technologies development. In this frame, suitable experiments were carried out on the CIRCE pool facility refurbished with the Integral Circulation Experiment (ICE) test section in order to investigate the thermal hydraulics and the heat transfer in grid spaced Fuel Pin Bundle cooled by liquid metal providing, among the others aim, experimental data in support of codes validation for the European fast reactor development. The study of thermal stratification in large pool reactor is relevant in the design of HLM nuclear reactor especially for safety issue. Thermal stratification should induce thermo-mechanical stresses on the structures and in accidental scenario conditions, could opposes to the establishment of natural circulation which is a fundamental aspect for the achievements of safety goals required in the GEN-IV roadmap. In the present work, a Protected Loss of Heat Sink with Loss Of Flow (PLOHS+LOF) scenario is experimentally simulated and the mixed convection with thermal stratification phenomena is investigated during the simulated transient, as foreseen in the frame of Horizon 2020 SESAME project [5]. A full characterization of thermal stratification inside the pool is presented, and the main results gained during the run are reported. The two tests named A (20 h) and B (6 h) here reported, essentially differs for the power supplied to the fuel bundle during the full power run (800 kW and 600 kW respectively). After the transition to natural circulation conditions, the power supplied to the bundle is decreased to about 30 kW simulating the decay heat. Finally the Nusselt number for the central subchannel of the fuel bundle simulator (FPS) is evaluated and compared with values obtained from Ushakov and Mikityuk correlations [6-7]. Copyright © 2016 by ASME.
9780791850046
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Utilizza questo identificativo per citare o creare un link a questo documento: http://hdl.handle.net/20.500.12079/3472
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