Development and testing of multicomponent fuel cladding with enhanced accidental performance

Kód výsledku v IS VaVaI

<a href="https://www.isvavai.cz/riv?ss=detail&h=RIV%2F26722445%3A_____%2F20%3AN0000020" target="_blank" >RIV/26722445:_____/20:N0000020 - isvavai.cz</a>

Nalezeny alternativní kódy

RIV/68407700:21220/20:00333943 RIV/68407700:21340/20:00333943

Výsledek na webu

<a href="https://www.journals.elsevier.com/nuclear-engineering-and-technology/" target="_blank" >https://www.journals.elsevier.com/nuclear-engineering-and-technology/</a>

DOI - Digital Object Identifier

<a href="http://dx.doi.org/10.1016/j.net.2019.08.015" target="_blank" >10.1016/j.net.2019.08.015</a>

Jazyk výsledku

angličtina

Název v původním jazyce

Development and testing of multicomponent fuel cladding with enhanced accidental performance

Popis výsledku v původním jazyce

Accident Tolerant Fuels have been widely studied since the Fukushima-Daiichi accident in 2011 as one of the options on how to further enhance the safety of nuclear power plants. Deposition of protective coatings on nuclear fuel claddings has been considered as a near-term concept that will reduce the high-temperature oxidation rate and enhance accidental tolerance of the cladding while providing additional benefits during normal operation and transients. This study focuses on experimental testing of Zr-based alloys coated with Cr-based coatings using Physical Vapour Deposition. The results of long-term corrosion tests, as well as tests simulating postulated accidents, are presented. Zr-1%Nb alloy used as nuclear fuel cladding serves as a substrate and Cr, CrN, CrxNy layers are deposited by unbalanced magnetron sputtering and reactive magnetron sputtering. The deposition procedures are optimized in order to improve coating properties. Coated as well as reference uncoated samples were experimentally tested. The presented results include standard long-term corrosion tests at 360 degrees C in WWER water chemistry, burst (creep) tests and mainly single and double-sided high-temperature steam oxidation tests between 1000 and 1400 degrees C related to postulated Loss-of-coolant accident and Design extension conditions. Coated and reference samples were characterized pre- and post-testing using mechanical testing (microhardness, ring compression test), Thermal Evolved Gas Analysis analysis (hydrogen, oxygen concentration), optical microscopy, scanning electron microscopy (EDS, WDS, EBSD) and X-ray diffraction.

Název v anglickém jazyce

Development and testing of multicomponent fuel cladding with enhanced accidental performance

Popis výsledku anglicky

Accident Tolerant Fuels have been widely studied since the Fukushima-Daiichi accident in 2011 as one of the options on how to further enhance the safety of nuclear power plants. Deposition of protective coatings on nuclear fuel claddings has been considered as a near-term concept that will reduce the high-temperature oxidation rate and enhance accidental tolerance of the cladding while providing additional benefits during normal operation and transients. This study focuses on experimental testing of Zr-based alloys coated with Cr-based coatings using Physical Vapour Deposition. The results of long-term corrosion tests, as well as tests simulating postulated accidents, are presented. Zr-1%Nb alloy used as nuclear fuel cladding serves as a substrate and Cr, CrN, CrxNy layers are deposited by unbalanced magnetron sputtering and reactive magnetron sputtering. The deposition procedures are optimized in order to improve coating properties. Coated as well as reference uncoated samples were experimentally tested. The presented results include standard long-term corrosion tests at 360 degrees C in WWER water chemistry, burst (creep) tests and mainly single and double-sided high-temperature steam oxidation tests between 1000 and 1400 degrees C related to postulated Loss-of-coolant accident and Design extension conditions. Coated and reference samples were characterized pre- and post-testing using mechanical testing (microhardness, ring compression test), Thermal Evolved Gas Analysis analysis (hydrogen, oxygen concentration), optical microscopy, scanning electron microscopy (EDS, WDS, EBSD) and X-ray diffraction.

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

Výsledek vznikl pri realizaci vícero projektů. Více informací v záložce Projekty.

Návaznosti

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

Rok uplatnění

2020

Kód důvěrnosti údajů

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

Název periodika

Nuclear Engineering and Technology

ISSN

1738-5733

e-ISSN

—

Svazek periodika

52

Číslo periodika v rámci svazku

3

Stát vydavatele periodika

KR - Korejská republika

Počet stran výsledku

13

Strana od-do

597-609

Kód UT WoS článku

000516803800017

EID výsledku v databázi Scopus

2-s2.0-85071492736