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

Result code in 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>

Result on the web

<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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Result language

angličtina

Original language name

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

Original language description

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.

Czech name

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Czech description

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Type

D - Article in proceedings

CEP classification

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OECD FORD branch

20501 - Materials engineering

Project

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Continuities

N - Vyzkumna aktivita podporovana z neverejnych zdroju

Publication year

2025

Confidentiality

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

Article name in the collection

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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Number of pages

10

Pages from-to

650-659

Publisher name

American Nuclear Society

Place of publication

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Event location

Nashville, USA

Event date

Oct 5, 2025

Type of event by nationality

WRD - Celosvětová akce

UT code for WoS article

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