In the context of the studies on GEN. IV/ADS nuclear systems, the correct evaluations of the temperature distribution in the fuel pin bundle is of central interest. In particular, the use of lead or lead-bismuth eutectic (LBE) as coolant for the new generation fast reactors is one of the most promising choices. Due to the high density and high conductivity of lead or LBE, a detailed analysis of the thermo-fluid dynamic behavior of the heavy liquid metal (HLM) inside the sub-channels of a fuel rod bundle is necessary in order to support the front-end engineering design (FEED) of GEN. IV/ADS prototypes and demonstrators. At ENEA-Brasimone R.C., large experimental facilities exist to study HLM free, forced and mixed convection in loops and pools: e.g. NACIE-UP is a large scale LBE loop for mixed convection experiments with a secondary side in pressurized water. In the context of the SEARCH FP7 project, an experiment was performed in the NACIE-UP facility to assess the coolability of a 19-pin wire-wrapped electrical bundle (Fuel Pin Simulator, FPS), with heat flux up to 1 MW/m2. The bundle is representative of the one adopted in the MYRRHA concept, and it is instrumented with 67 thermocouples to monitor bulk and wall temperatures in the different ranks of subchannels at different heights. The mass flow rate in the loop is measured by an induction flow meter and by thermal balance in stationary conditions. A large experimental test matrix has to be carried out in order to characterize the coolability of MYRRHA Fuel Sub-Assembly. Results in terms of wall and bulk temperatures, heat transfer coefficients and pin azimuthal temperature distribution are available at different mass flow rates in the range 0-7 kg/s, i.e. up to 1 m/s in the subchannel, half of the MYRRHA FA nominal velocity. The first experimental tests with FPS are provided in the paper. Those are the first data on HLM-cooled wire-wrapped Fuel Assembly reproducing the MYRRHA FA. Post-test analysis and error propagation study on derived quantities is also provided. Copyright © (2015) by American Nuclear Society All rights reserved.

Experimental fuel pin bundle characterization in the NACIE-UP HLM facility

Tarantino, M.;Laffi, L.;Sermenghi, V.;Polazzi, G.;Di Piazza, I.
2015

Abstract

In the context of the studies on GEN. IV/ADS nuclear systems, the correct evaluations of the temperature distribution in the fuel pin bundle is of central interest. In particular, the use of lead or lead-bismuth eutectic (LBE) as coolant for the new generation fast reactors is one of the most promising choices. Due to the high density and high conductivity of lead or LBE, a detailed analysis of the thermo-fluid dynamic behavior of the heavy liquid metal (HLM) inside the sub-channels of a fuel rod bundle is necessary in order to support the front-end engineering design (FEED) of GEN. IV/ADS prototypes and demonstrators. At ENEA-Brasimone R.C., large experimental facilities exist to study HLM free, forced and mixed convection in loops and pools: e.g. NACIE-UP is a large scale LBE loop for mixed convection experiments with a secondary side in pressurized water. In the context of the SEARCH FP7 project, an experiment was performed in the NACIE-UP facility to assess the coolability of a 19-pin wire-wrapped electrical bundle (Fuel Pin Simulator, FPS), with heat flux up to 1 MW/m2. The bundle is representative of the one adopted in the MYRRHA concept, and it is instrumented with 67 thermocouples to monitor bulk and wall temperatures in the different ranks of subchannels at different heights. The mass flow rate in the loop is measured by an induction flow meter and by thermal balance in stationary conditions. A large experimental test matrix has to be carried out in order to characterize the coolability of MYRRHA Fuel Sub-Assembly. Results in terms of wall and bulk temperatures, heat transfer coefficients and pin azimuthal temperature distribution are available at different mass flow rates in the range 0-7 kg/s, i.e. up to 1 m/s in the subchannel, half of the MYRRHA FA nominal velocity. The first experimental tests with FPS are provided in the paper. Those are the first data on HLM-cooled wire-wrapped Fuel Assembly reproducing the MYRRHA FA. Post-test analysis and error propagation study on derived quantities is also provided. Copyright © (2015) by American Nuclear Society All rights reserved.
9781510811843
Thermal-hydraulics;Fast reactor;Liquid metal
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Utilizza questo identificativo per citare o creare un link a questo documento: https://hdl.handle.net/20.500.12079/3863
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