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

Kód výsledku v 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>

Výsledek na webu

<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>

Jazyk výsledku

angličtina

Název v původním jazyce

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

Popis výsledku v původním jazyce

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.

Název v anglickém jazyce

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

Popis výsledku anglicky

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.

Druh

J_imp - Článek v periodiku v databázi Web of Science

CEP obor

—

OECD FORD obor

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

Projekt

—

Návaznosti

I - Institucionalni podpora na dlouhodoby koncepcni rozvoj vyzkumne organizace

Rok uplatnění

2021

Kód důvěrnosti údajů

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

Název periodika

Nuclear Engineering and Design

ISSN

0029-5493

e-ISSN

1872-759X

Svazek periodika

376

Číslo periodika v rámci svazku

May

Stát vydavatele periodika

CH - Švýcarská konfederace

Počet stran výsledku

16

Strana od-do

1-16

Kód UT WoS článku

000641694500003

EID výsledku v databázi Scopus

2-s2.0-85101658382