NEAMS IRP Challenge Problem 1: Flexible Modeling for Heat Transfer for Applications in Advanced Reactors

Igor A. Bolotnov, Arsen S. Iskhakov, Tri Nguyen, Cheng Kai Tai, Ralph Wiser, Emilio Baglietto, Nam Dinh, Dillon Shaver, Elia Merzari

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

2 Scopus citations

Abstract

The adoption of liquid metals and molten salts as coolant fluids in advanced reactor designs has challenged traditional turbulence and heat transfer models because of vastly different heat transfer characteristics from water. The challenge is exacerbated by the yet-to-be-understood mixed convection regime, which is crucial to passive heat removal. The NEAMS IRP Challenge Problem 1 (CP1) aims to facilitate scale-flexible heat transfer models that can serve for broader applications in reactor thermal-hydraulic analysis. CP1 research activities include direct numerical simulation (DNS) and the development of advanced turbulence modeling techniques. The DNS efforts studies the low- and high-Prandtl mixed convection in the identified canonical flow scenario, from which the fundamental understanding on the effect of buoyancy on turbulence and heat transfer is investigated. The CP1 modeling endeavors aim at advancement of the present engineering models and the employment of the data driven (DD) methods for the turbulence modeling. In the engineering model concentration, the performance of the current CFD models is assessed in the non-unitary Prandtl flows to gain physical understanding of their shortcomings. A framework for model form error prediction and error propagation estimation is also established. A DD RANS framework is developed for the frozen prediction of Reynolds stress and turbulent heat flux based on the polynomial tensor representation of the respective quantities. The DD framework showed satisfactory performance over the k − τ model in the forced and mixed convection cases.

Original languageEnglish (US)
Title of host publicationProceedings of the 20th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2023
PublisherAmerican Nuclear Society
Pages4450-4463
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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