In the context of GEN IV innovative nuclear reactor design, Lead Fast Reactor (LFR) represents a promising alternative to Sodium Fast Reactor (SFR). European Research Community is promoting R&D programs to guide the design of Pilot Plant for the advancement of lead fast reactor technology. For this purpose, ENEA Brasimone R.C. is conducting several thermal-hydraulic experimental campaigns aiming at supporting the design of the Multipurpose Hybrid Research Reactor for High-tech Application (MYRRHA). ENEA is involved in fuel pin & assembly thermal-hydraulic investigations through experimental activities on the NACIE loop type and CIRCE pool type facilities. The aim of this work, performed in the frame of the partnership ENEA-Pisa University, is the investigation of heat transfer phenomena in fuel rod bundles in a pool type configuration. The obtained experimental data represent the first data published for LBE coolant in a pool type configuration. The experimental campaign is conducted on the CIRCE-ICE (Circulation Eutectic in the Integral Circulation Experiment configuration) facility, filled with 70 tons of molten LBE. A series of experiments performed in natural and forced circulation regimes are carried out and obtained data are post-processed obtaining a Nusselt characterization in the central subchannel of the bundle as function of the Peclet number in a Peclet range of 500-3000. Experimentally determined values, are then compared with correlations available in the HLM literature (obtained using NaK or Hg as working fluid), showing a linear trend as the Pe increases, in agreement with the proposed correlations, showing a general tendency to lie below them. In particular, the Nu numbers obtained from the experimental data are about 25% lower than data predicted from correlations by less than 25%. Moreover, the paper reports a preliminary analysis and discussion of such results, also in comparison with CFD calculations performed by CFX code. Copyright © (2015) by American Nuclear Society All rights reserved.
Fuel pin bundle experimental characterization in HLM large pool system
Di Piazza, I.;Tarantino, M.;Agostini, P.
2015-01-01
Abstract
In the context of GEN IV innovative nuclear reactor design, Lead Fast Reactor (LFR) represents a promising alternative to Sodium Fast Reactor (SFR). European Research Community is promoting R&D programs to guide the design of Pilot Plant for the advancement of lead fast reactor technology. For this purpose, ENEA Brasimone R.C. is conducting several thermal-hydraulic experimental campaigns aiming at supporting the design of the Multipurpose Hybrid Research Reactor for High-tech Application (MYRRHA). ENEA is involved in fuel pin & assembly thermal-hydraulic investigations through experimental activities on the NACIE loop type and CIRCE pool type facilities. The aim of this work, performed in the frame of the partnership ENEA-Pisa University, is the investigation of heat transfer phenomena in fuel rod bundles in a pool type configuration. The obtained experimental data represent the first data published for LBE coolant in a pool type configuration. The experimental campaign is conducted on the CIRCE-ICE (Circulation Eutectic in the Integral Circulation Experiment configuration) facility, filled with 70 tons of molten LBE. A series of experiments performed in natural and forced circulation regimes are carried out and obtained data are post-processed obtaining a Nusselt characterization in the central subchannel of the bundle as function of the Peclet number in a Peclet range of 500-3000. Experimentally determined values, are then compared with correlations available in the HLM literature (obtained using NaK or Hg as working fluid), showing a linear trend as the Pe increases, in agreement with the proposed correlations, showing a general tendency to lie below them. In particular, the Nu numbers obtained from the experimental data are about 25% lower than data predicted from correlations by less than 25%. Moreover, the paper reports a preliminary analysis and discussion of such results, also in comparison with CFD calculations performed by CFX code. Copyright © (2015) by American Nuclear Society All rights reserved.I documenti in IRIS sono protetti da copyright e tutti i diritti sono riservati, salvo diversa indicazione.
