Microstructure and Nanoindentation Studies of Hydridated Zircaloy-4 Claddings After High Temperature Oxidation

Result code in IS VaVaI

<a href="https://www.isvavai.cz/riv?ss=detail&h=RIV%2F26722445%3A_____%2F21%3AN0000047" target="_blank" >RIV/26722445:_____/21:N0000047 - isvavai.cz</a>

Result on the web

<a href="https://asmedigitalcollection.asme.org/nuclearengineering/article-abstract/7/2/021604/1090566/Microstructure-and-Nanoindentation-Studies-of?redirectedFrom=fulltext" target="_blank" >https://asmedigitalcollection.asme.org/nuclearengineering/article-abstract/7/2/021604/1090566/Microstructure-and-Nanoindentation-Studies-of?redirectedFrom=fulltext</a>

DOI - Digital Object Identifier

<a href="http://dx.doi.org/10.1115/1.4049053" target="_blank" >10.1115/1.4049053</a>

Result language

angličtina

Original language name

Microstructure and Nanoindentation Studies of Hydridated Zircaloy-4 Claddings After High Temperature Oxidation

Original language description

Zirconium-based alloys are one of the most significant materials in thermal-neutron reactor systems. With very low neutron capture cross section, good corrosion resistance, mechanical strength and resistance to neutron radiation damage, zirconium alloys are used for fuel claddings. Cladding materials are still improved and tested in normal as well as critical reactor conditions. Zircaloy-4 (Zr-1.5Sn-0.2Fe-0.1Cr) is used for west types of light-water reactors, Pressurized Water Reactors (PWR). In our study, Zircaloy-4 cladding tubes were high-temperature oxidized in steam at the series of temperatures from 950 up to 1425 °C to simulate PWR reaching severe accident conditions. To observe the influence of hydrogen (H) diffusing from the coolant water on oxidation process, the specimens with ∼1000 ppm H were compared to the specimens with almost no hydrogen content. Wave Dispersive Spectroscopy (WDS) and nanoindentation were performed in line profiles across the cladding wall. Both methods contributed to verify the pseudobinary Zircaloy-4/oxygen phase diagram with focus on determination of phase boundaries. The increase of oxygen concentration with increasing temperature was observed. Moreover, oxygen concentration profiles and related change in nanohardness and Young's modulus showed the effect of hydrogen on the cladding microstructure. Hydrogen dissolved in metallic matrix increases the oxygen solubility in prior β-phase, the specimens with 1000 ppm H showed the higher oxygen content at almost all temperatures. As well, material hardening was observed on specimens with 1000 ppm H with significant difference in β-phase, measured on specimens exposed to lowest and highest oxidation temperature. Thus, with increasing temperature and hydrogen content, increased oxygen solubility affects the cladding ductility.

Czech name

—

Czech description

—

Type

J_imp - Article in a specialist periodical, which is included in the Web of Science database

CEP classification

—

OECD FORD branch

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

Project

<a href="/en/project/LQ1603" target="_blank" >LQ1603: Research for SUSEN</a><br>

Continuities

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

Publication year

2021

Confidentiality

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

Name of the periodical

Journal of Nuclear Engineering and Radiation Science

ISSN

2332-8983

e-ISSN

2332-8975

Volume of the periodical

7

Issue of the periodical within the volume

2

Country of publishing house

NL - THE KINGDOM OF THE NETHERLANDS

Number of pages

6

Pages from-to

1-6

UT code for WoS article

000630005800022

EID of the result in the Scopus database

2-s2.0-85098913292