LOFA and LOUHS analysis on the energy well SMR using TRACE code

Kód výsledku v IS VaVaI

<a href="https://www.isvavai.cz/riv?ss=detail&h=RIV%2F68407700%3A21220%2F24%3A00371706" target="_blank" >RIV/68407700:21220/24:00371706 - isvavai.cz</a>

Výsledek na webu

<a href="https://doi.org/10.1016/j.pnucene.2024.105054" target="_blank" >https://doi.org/10.1016/j.pnucene.2024.105054</a>

DOI - Digital Object Identifier

<a href="http://dx.doi.org/10.1016/j.pnucene.2024.105054" target="_blank" >10.1016/j.pnucene.2024.105054</a>

Jazyk výsledku

angličtina

Název v původním jazyce

LOFA and LOUHS analysis on the energy well SMR using TRACE code

Popis výsledku v původním jazyce

The removal of decay heat is the key to the management of nuclear reactor accidents. For a concept of a small modular reactor Energy Well (EW), some of the most severe possible accidents would be loss of flow (LOFA) and also LOFA with the loss of ultimate heat sink (LOUHS). EW is a fluoride salt-cooled high-temperature reactor (FHR) with TRISO microparticles fuel in carbon matrix, molten salt FLiBe as a coolant and a graphite moderator. Some of its key features are a seven-year fuel cycle with fuel exchange off-site, easy transportability, and power plant modularity, all of which makes it an interesting contender among the energy sources of the near future. In order to assess the temperatures in key points of the EW primary circuit during the aforementioned accident scenarios, the EW model in thermohydraulic code TRACE was adjusted and scenarios with the trip of one through six pumps were modelled for both LOFA and LOFA with LOUHS. The point was to investigate whether any of these scenarios would cause failure of the primary circuit components, such as the fuel cladding, reactor vessel, or the surrounding concrete. None of the safety criteria were exceeded.

Název v anglickém jazyce

LOFA and LOUHS analysis on the energy well SMR using TRACE code

Popis výsledku anglicky

The removal of decay heat is the key to the management of nuclear reactor accidents. For a concept of a small modular reactor Energy Well (EW), some of the most severe possible accidents would be loss of flow (LOFA) and also LOFA with the loss of ultimate heat sink (LOUHS). EW is a fluoride salt-cooled high-temperature reactor (FHR) with TRISO microparticles fuel in carbon matrix, molten salt FLiBe as a coolant and a graphite moderator. Some of its key features are a seven-year fuel cycle with fuel exchange off-site, easy transportability, and power plant modularity, all of which makes it an interesting contender among the energy sources of the near future. In order to assess the temperatures in key points of the EW primary circuit during the aforementioned accident scenarios, the EW model in thermohydraulic code TRACE was adjusted and scenarios with the trip of one through six pumps were modelled for both LOFA and LOFA with LOUHS. The point was to investigate whether any of these scenarios would cause failure of the primary circuit components, such as the fuel cladding, reactor vessel, or the surrounding concrete. None of the safety criteria were exceeded.

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

2024

Kód důvěrnosti údajů

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

Název periodika

Progress in Nuclear Energy

ISSN

0149-1970

e-ISSN

1878-4224

Svazek periodika

169

Číslo periodika v rámci svazku

4

Stát vydavatele periodika

US - Spojené státy americké

Počet stran výsledku

7

Strana od-do

1-7

Kód UT WoS článku

001169935000001

EID výsledku v databázi Scopus

2-s2.0-85184779117