Experimental investigation and predictive modeling of the thermal creep behavior of Zr1Nb alloy cladding

Kód výsledku v IS VaVaI

<a href="https://www.isvavai.cz/riv?ss=detail&h=RIV%2F68081723%3A_____%2F25%3A00638107" target="_blank" >RIV/68081723:_____/25:00638107 - isvavai.cz</a>

Výsledek na webu

<a href="https://link.springer.com/article/10.1007/s10853-025-11268-9" target="_blank" >https://link.springer.com/article/10.1007/s10853-025-11268-9</a>

DOI - Digital Object Identifier

<a href="http://dx.doi.org/10.1007/s10853-025-11268-9" target="_blank" >10.1007/s10853-025-11268-9</a>

Jazyk výsledku

angličtina

Název v původním jazyce

Experimental investigation and predictive modeling of the thermal creep behavior of Zr1Nb alloy cladding

Popis výsledku v původním jazyce

Zirconium alloys are used as a fuel cladding material in light water reactors due to their low neutron cross section, good corrosion resistance, appropriate high-temperature strength, and dimensional stability under radiation. Thermal creep is considered one of the severe degradation factors during the reactor core operation. Therefore, an understanding of the creep flow processes and fracture control is necessary for predicting the lifetime of the cladding component. This paper summarizes the results from an extensive experiment on the characteristics of creep flow of the non-irradiated Zr1%Nb cladding alloy in the form of a thin-walled cladding tube. The study is based on the experimental results of the constant stress creep tests and the sophisticated helicoid spring specimen technique. Finally, the modified creep rate model has been proposed, where the driving force is expressed as effective stress instead of the applied stress. The effective stress was described as linear dependence on the applied stress with some small residual stress below the threshold.

Název v anglickém jazyce

Experimental investigation and predictive modeling of the thermal creep behavior of Zr1Nb alloy cladding

Popis výsledku anglicky

Zirconium alloys are used as a fuel cladding material in light water reactors due to their low neutron cross section, good corrosion resistance, appropriate high-temperature strength, and dimensional stability under radiation. Thermal creep is considered one of the severe degradation factors during the reactor core operation. Therefore, an understanding of the creep flow processes and fracture control is necessary for predicting the lifetime of the cladding component. This paper summarizes the results from an extensive experiment on the characteristics of creep flow of the non-irradiated Zr1%Nb cladding alloy in the form of a thin-walled cladding tube. The study is based on the experimental results of the constant stress creep tests and the sophisticated helicoid spring specimen technique. Finally, the modified creep rate model has been proposed, where the driving force is expressed as effective stress instead of the applied stress. The effective stress was described as linear dependence on the applied stress with some small residual stress below the threshold.

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

<a href="/cs/project/TN02000018" target="_blank" >TN02000018: Národní Centrum Kompetence STROJÍRENSTVÍ</a><br>

Návaznosti

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

Rok uplatnění

2025

Kód důvěrnosti údajů

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

Název periodika

Journal of Materials Science

ISSN

0022-2461

e-ISSN

1573-4803

Svazek periodika

60

Číslo periodika v rámci svazku

39

Stát vydavatele periodika

US - Spojené státy americké

Počet stran výsledku

7

Strana od-do

18352-18358

Kód UT WoS článku

001545108000001

EID výsledku v databázi Scopus

2-s2.0-105012752799