EXPERIMENTAL BEHAVIOR OF CHROMIUM-BASED COATINGS

Kód výsledku v IS VaVaI

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Nalezeny alternativní kódy

RIV/68407700:21340/18:00337524

Výsledek na webu

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DOI - Digital Object Identifier

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Jazyk výsledku

angličtina

Název v původním jazyce

EXPERIMENTAL BEHAVIOR OF CHROMIUM-BASED COATINGS

Popis výsledku v původním jazyce

Accident Tolerant Fuels have been widely studied since the Fukushima accident in 2011. Deposition of protective coatings on nuclear fuel claddings has been considered as a near-term concept that will reduce the high-temperature oxidation rate of zirconium-based alloys and enhance accident tolerance of reactor cores by providing additional coping time. This study is focused on low-temperature oxidation and LOCA experimental behavior of Zr-1%Nb cladding alloys coated with Cr and CrN based layers by PVD technique. Coated and reference uncoated samples of Zr-1%Nb alloy were tested in several experiments. The presented results include standard corrosion tests at 360°C in WWER water chemistry, burst tests and mainly single and double-sided high-temperature steam oxidation between 1000 and 1400°C. Coated and reference samples were characterized pre- and post-testing using mechanical testing (microhardness, ring compression test), TEA analysis (hydrogen, oxygen measurements), optical microscopy, scanning electron microscopy (EDS, WDS, EBSD) and other techniques including X-ray diffraction.

Název v anglickém jazyce

EXPERIMENTAL BEHAVIOR OF CHROMIUM-BASED COATINGS

Popis výsledku anglicky

Accident Tolerant Fuels have been widely studied since the Fukushima accident in 2011. Deposition of protective coatings on nuclear fuel claddings has been considered as a near-term concept that will reduce the high-temperature oxidation rate of zirconium-based alloys and enhance accident tolerance of reactor cores by providing additional coping time. This study is focused on low-temperature oxidation and LOCA experimental behavior of Zr-1%Nb cladding alloys coated with Cr and CrN based layers by PVD technique. Coated and reference uncoated samples of Zr-1%Nb alloy were tested in several experiments. The presented results include standard corrosion tests at 360°C in WWER water chemistry, burst tests and mainly single and double-sided high-temperature steam oxidation between 1000 and 1400°C. Coated and reference samples were characterized pre- and post-testing using mechanical testing (microhardness, ring compression test), TEA analysis (hydrogen, oxygen measurements), optical microscopy, scanning electron microscopy (EDS, WDS, EBSD) and other techniques including X-ray diffraction.

Druh

D - Stať ve sborníku

CEP obor

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OECD FORD obor

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

Projekt

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Návaznosti

V - Vyzkumna aktivita podporovana z jinych verejnych zdroju

Rok uplatnění

2018

Kód důvěrnosti údajů

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

Název statě ve sborníku

TopFuel 2018 Water Reactor Fuel Performance

ISBN

978-92-95064-32-4

ISSN

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e-ISSN

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Počet stran výsledku

13

Strana od-do

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Název nakladatele

European Nuclear Society

Místo vydání

Brusel

Místo konání akce

Praha

Datum konání akce

30. 9. 2018

Typ akce podle státní příslušnosti

WRD - Celosvětová akce

Kód UT WoS článku

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