Computational fluid dynamics modelling of lead natural convection and solidification in a pool type geometry

Result code in IS VaVaI

<a href="https://www.isvavai.cz/riv?ss=detail&h=RIV%2F26722445%3A_____%2F21%3AN0000017" target="_blank" >RIV/26722445:_____/21:N0000017 - isvavai.cz</a>

Result on the web

<a href="https://www.sciencedirect.com/science/article/pii/S002954932100056X" target="_blank" >https://www.sciencedirect.com/science/article/pii/S002954932100056X</a>

DOI - Digital Object Identifier

<a href="http://dx.doi.org/10.1016/j.nucengdes.2021.111104" target="_blank" >10.1016/j.nucengdes.2021.111104</a>

Result language

angličtina

Original language name

Computational fluid dynamics modelling of lead natural convection and solidification in a pool type geometry

Original language description

Lead-cooled fast reactors (LFRs) are being studied by the organisations in the Generation IV International Forum (GIF) due to molten lead?s good thermodynamic properties, nuclear sustainability and safety. The study of lead solidification in a lead-cooled fast reactor is critical for the safety analysis of the reactor. Lead freezing may lead to overheating of the fuel assemblies or other components in the primary circuits. An activity that is focused on the development of the numerical models that deal with lead thermal hydraulics and solidification was ongoing within the H2020 project SESAME. The computational activity was supported by an experimental campaign. The SESAME stand experimental facility was assembled and operated at the Research Centre Rez (CVR) for the collection of thermal-hydraulic data on lead natural convection and solidification in a vessel type geometry. Simultaneously, two computational fluid dynamics (CFD) models of the SESAME stand were developed using ANSYS FLUENT and STAR-CCM + software. The models are benchmarked against the experimental data for both the steady-state and transient regimes. The methodology of the ANSYS FLUENT model has been described in detail, and the results were compared with both the experimental data and the STAR-CCM + model. The capability of the numerical model to deal with the lead thermal hydraulic phenomena and their shortcomings is discussed. The challenges and the lessons learned from both the experimental and numerical activities are presented to support the development of computational tools for the lead-cooled nuclear reactors and their safety assessment.

Czech name

—

Czech description

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Type

J_imp - Article in a specialist periodical, which is included in the Web of Science database

CEP classification

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OECD FORD branch

20305 - Nuclear related engineering; (nuclear physics to be 1.3);

Project

—

Continuities

I - Institucionalni podpora na dlouhodoby koncepcni rozvoj vyzkumne organizace

Publication year

2021

Confidentiality

S - Úplné a pravdivé údaje o projektu nepodléhají ochraně podle zvláštních právních předpisů

Name of the periodical

Nuclear Engineering and Design

ISSN

0029-5493

e-ISSN

1872-759X

Volume of the periodical

376

Issue of the periodical within the volume

May

Country of publishing house

CH - SWITZERLAND

Number of pages

16

Pages from-to

1-16

UT code for WoS article

000641694500003

EID of the result in the Scopus database

2-s2.0-85101658382