COUPLED THERMAL-HYDRAULICS AND NEUTRON TRANSPORT CALCULATIONS OF SMALL MODULAR REACTOR USING SERPENT, OPENFOAM AND SUBCHANFLOW CODES

Kód výsledku v IS VaVaI

<a href="https://www.isvavai.cz/riv?ss=detail&h=RIV%2F68407700%3A21730%2F24%3A00381178" target="_blank" >RIV/68407700:21730/24:00381178 - isvavai.cz</a>

Výsledek na webu

<a href="https://doi.org/10.1115/ICONE31-136127" target="_blank" >https://doi.org/10.1115/ICONE31-136127</a>

DOI - Digital Object Identifier

<a href="http://dx.doi.org/10.1115/ICONE31-136127" target="_blank" >10.1115/ICONE31-136127</a>

Jazyk výsledku

angličtina

Název v původním jazyce

COUPLED THERMAL-HYDRAULICS AND NEUTRON TRANSPORT CALCULATIONS OF SMALL MODULAR REACTOR USING SERPENT, OPENFOAM AND SUBCHANFLOW CODES

Popis výsledku v původním jazyce

Nowadays, several Small Modular Reactors are under development and high-fidelity neutron transport calculations with thermal-hydraulic feedback are a powerful tool for design and optimization of such reactors. The present work focuses on analyses of heavy water small modular reactor using coupled CFD, Monte Carlo and subchannel code. The coupling of Monte Carlo and subchannel codes was based on Picard iterations with power relaxation using stochastic approach. CFD simulations served for a generation of inflow conditions for subchannel code. The coupled pin-by-pin subchannel analyses was compared against fuel assembly level subchannel analyses. Additionally, a case with asymmetrical inflow conditions was analyses using coupled codes to reveal influence of inflow uncertainty on criticality and maximum fuel temperatures. Comparison of coupled calculations on pin-by-pin level and fuel assembly level showed that a lower resolution led to underprediction of maximum fuel temperatures by 108 °C and maximum cladding temperature by 9 °C. Further, the low resolution led to the underprediction of multiplication factor by 410 pcm. The asymmetrical inflow case led to a slightly higher coolant, cladding, and fuel temperatures in order of few degrees due to a lower flow rate in half of the core. Further, the asymmetric condition did not influence the reactivity as the multiplication factor changed insignificantly.

Název v anglickém jazyce

COUPLED THERMAL-HYDRAULICS AND NEUTRON TRANSPORT CALCULATIONS OF SMALL MODULAR REACTOR USING SERPENT, OPENFOAM AND SUBCHANFLOW CODES

Popis výsledku anglicky

Nowadays, several Small Modular Reactors are under development and high-fidelity neutron transport calculations with thermal-hydraulic feedback are a powerful tool for design and optimization of such reactors. The present work focuses on analyses of heavy water small modular reactor using coupled CFD, Monte Carlo and subchannel code. The coupling of Monte Carlo and subchannel codes was based on Picard iterations with power relaxation using stochastic approach. CFD simulations served for a generation of inflow conditions for subchannel code. The coupled pin-by-pin subchannel analyses was compared against fuel assembly level subchannel analyses. Additionally, a case with asymmetrical inflow conditions was analyses using coupled codes to reveal influence of inflow uncertainty on criticality and maximum fuel temperatures. Comparison of coupled calculations on pin-by-pin level and fuel assembly level showed that a lower resolution led to underprediction of maximum fuel temperatures by 108 °C and maximum cladding temperature by 9 °C. Further, the low resolution led to the underprediction of multiplication factor by 410 pcm. The asymmetrical inflow case led to a slightly higher coolant, cladding, and fuel temperatures in order of few degrees due to a lower flow rate in half of the core. Further, the asymmetric condition did not influence the reactivity as the multiplication factor changed insignificantly.

Druh

D - Stať ve sborníku

CEP obor

—

OECD FORD obor

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

Projekt

<a href="/cs/project/TN02000012" target="_blank" >TN02000012: Centrum pokročilých jaderných technologií II</a><br>

Návaznosti

P - Projekt vyzkumu a vyvoje financovany z verejnych zdroju (s odkazem do CEP)

Rok uplatnění

2024

Kód důvěrnosti údajů

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

Název statě ve sborníku

Proceedings of 2024 31st International Conference on Nuclear Engineering, ICONE 2024

ISBN

978-0-7918-8827-8

ISSN

—

e-ISSN

—

Počet stran výsledku

6

Strana od-do

—

Název nakladatele

American Society of Mechanical Engineers - ASME

Místo vydání

New York

Místo konání akce

Praha

Datum konání akce

4. 8. 2024

Typ akce podle státní příslušnosti

WRD - Celosvětová akce

Kód UT WoS článku

001349529200049