Evolution of Microstructure of Advanced Zirconium Alloys with Increasing Level of Irradiation

Kód výsledku v IS VaVaI

<a href="https://www.isvavai.cz/riv?ss=detail&h=RIV%2F26722445%3A_____%2F25%3AN0000151" target="_blank" >RIV/26722445:_____/25:N0000151 - isvavai.cz</a>

Výsledek na webu

<a href="https://www.ans.org/pubs/proceedings/article-59438/" target="_blank" >https://www.ans.org/pubs/proceedings/article-59438/</a>

DOI - Digital Object Identifier

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Jazyk výsledku

angličtina

Název v původním jazyce

Evolution of Microstructure of Advanced Zirconium Alloys with Increasing Level of Irradiation

Popis výsledku v původním jazyce

Continuous development of zirconium alloys based fuel claddings as a second barrier against leakage of fission products into the environment requires adequate study of their material property evolution and changes. These changes are very important not only from the point of view of the material performance during its operation in the reactor, but also from the point of view of the fuel long-term storage in the spent fuel pool or dry casks after irradiation in the reactor. A large experimental and research program focused on irradiation induced growth and microstructural changes of Zr-alloys with different composition and heat treatment is performed by ALVEL in cooperation with other companies. Dedicated Material Cluster Assemblies (MCAs) with samples of advanced zirconium alloys have been irradiated in the reactor core of VVER-1000 reactor for up to five cycles followed by PIE. The work is scheduled to be finished in 2026. This paper is focused on radiation-induced growth and microstructural changes of these samples with increasing accumulated neutron fluence. The results of samples after one, two, three and four irradiation cycles compared with irradiation induced growth and accumulated neutron fluence are preseted in the paper. Detailed attention is paid to the study of radiation-induced precipitates, hydrides and a- and c-type dislocation loops. The results of the whole project provide a comprehensive picture of the material properties and expected nuclear fuel cladding and guiding tubes behaviour during its lifetime and how it is affected by the heat treatment at the end of the manufacturing process.

Název v anglickém jazyce

Evolution of Microstructure of Advanced Zirconium Alloys with Increasing Level of Irradiation

Popis výsledku anglicky

Continuous development of zirconium alloys based fuel claddings as a second barrier against leakage of fission products into the environment requires adequate study of their material property evolution and changes. These changes are very important not only from the point of view of the material performance during its operation in the reactor, but also from the point of view of the fuel long-term storage in the spent fuel pool or dry casks after irradiation in the reactor. A large experimental and research program focused on irradiation induced growth and microstructural changes of Zr-alloys with different composition and heat treatment is performed by ALVEL in cooperation with other companies. Dedicated Material Cluster Assemblies (MCAs) with samples of advanced zirconium alloys have been irradiated in the reactor core of VVER-1000 reactor for up to five cycles followed by PIE. The work is scheduled to be finished in 2026. This paper is focused on radiation-induced growth and microstructural changes of these samples with increasing accumulated neutron fluence. The results of samples after one, two, three and four irradiation cycles compared with irradiation induced growth and accumulated neutron fluence are preseted in the paper. Detailed attention is paid to the study of radiation-induced precipitates, hydrides and a- and c-type dislocation loops. The results of the whole project provide a comprehensive picture of the material properties and expected nuclear fuel cladding and guiding tubes behaviour during its lifetime and how it is affected by the heat treatment at the end of the manufacturing process.

Druh

D - Stať ve sborníku

CEP obor

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OECD FORD obor

20501 - Materials engineering

Projekt

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Návaznosti

N - Vyzkumna aktivita podporovana z neverejnych zdroju

Rok uplatnění

2025

Kód důvěrnosti údajů

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

Název statě ve sborníku

Proceedings of the TopFuel 2025: Nuclear Reactor Fuel Performance Conference

ISBN

978-089448228-1

ISSN

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e-ISSN

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Počet stran výsledku

10

Strana od-do

650-659

Název nakladatele

American Nuclear Society

Místo vydání

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Místo konání akce

Nashville, USA

Datum konání akce

5. 10. 2025

Typ akce podle státní příslušnosti

WRD - Celosvětová akce

Kód UT WoS článku

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