Code Validation of SAM Using Forced and Natural-Circulation Data from NACIE-UP Benchmark

Victor Coppo Leite, Elia Merzari, Adam Dix, Ling Zou

Research output: Chapter in Book/Report/Conference proceedingConference contribution

2 Scopus citations

Abstract

Heavy liquid metals (HLMs) are promising candidates as coolants of GEN-IV fast reactors due to their thermo-physical properties. In the last decade, experimental works have been proposed as part of R&D efforts to develop advanced reactors. In this context, the NACIE-UP facility at the ENEA Brasimone Research Centre (Italy) has conducted many experiments to study the thermo-fluid dynamic behavior of HLM in rod bundle configurations. In the present work, SAM, a modern system analysis code developed at Argonne National Laboratory, is validated with experiments from the NACIE-UP facility. This facility consists of a rectangular loop operated with lead-bismuth eutectic (LBE). A wire-spaced 19-pin fuel bundle simulator is installed on the bottom of the riser portion of the experimental loop, which is equipped with a set of sensors to monitor relevant parameters, i.e., temperatures, heat transfer, and flow conditions (mass flow rate) at different locations. Both natural circulation and forced convection cases were addressed. The SAM model developed in the present work agrees with the experimental facility. Relative errors for temperatures and mass flow rates are within the uncertainty of closure models used in the SAM compared to experimental results.

Original languageEnglish (US)
Title of host publicationProceedings of the 20th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2023
PublisherAmerican Nuclear Society
Pages296-309
Number of pages14
ISBN (Electronic)9780894487934
DOIs
StatePublished - 2023
Event20th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2023 - Washington, United States
Duration: Aug 20 2023Aug 25 2023

Publication series

NameProceedings of the 20th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2023

Conference

Conference20th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2023
Country/TerritoryUnited States
CityWashington
Period8/20/238/25/23

All Science Journal Classification (ASJC) codes

  • Nuclear Energy and Engineering
  • Instrumentation

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