Local mechanical properties testing of zirconium alloy nuclear fuel claddings

Kód výsledku v IS VaVaI

<a href="https://www.isvavai.cz/riv?ss=detail&h=RIV%2F26722445%3A_____%2F22%3AN0000230" target="_blank" >RIV/26722445:_____/22:N0000230 - isvavai.cz</a>

Výsledek na webu

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DOI - Digital Object Identifier

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Jazyk výsledku

angličtina

Název v původním jazyce

Local mechanical properties testing of zirconium alloy nuclear fuel claddings

Popis výsledku v původním jazyce

Zirconium alloys are being commonly used as a material of choice for nuclear fuel claddings in water cooled nuclear reactors for decades due to their good corrosion resistance and low neutron absorption. However, the increasing operation conditions of next generation nuclear reactors (Gen-IV) in terms of higher temperatures, pressures and higher neutron flux requires evaluation of further Zr cladding usability. The embrittlement of Zr claddings due to hydrogen pickup from reactor coolant is one of the issues for its potential use in Gen-IV reactors. Nanoindentation is an effective tool for analysis of the change of mechanical properties of hydrogen enriched Zr claddings from localised material volume. Zirconium alloy Zr-1Nb (E110) with experimentally induced hydrides was analysed by the means of nanoindentation. Zirconium hydrides were formed in the material after exposure in high temperature water autoclave. The nanoindentation measurements were performed as an array of 10x10 indents across areas with hydrides. Depth dependent mechanical properties measured by nanoindentation were compared between the material with no hydrogen content, low hydrogen content (127 ppm H) and high hydrogen content (397 ppm H). Another set of nanoindentation measurements were performed on Zircalloy-4 (Zr-1.5Sn-0.2Fe-0.1Cr) cladding tubes after high-temperature oxidation (950°C – 1425°C) in steam, simulating severe accident conditions in pressurised water reactors (PWRs). Linear nanoindentation profile measurements were conducted in radial direction of the cladding across ZrO2 layer, oxygen enriched α-Zr phase, α-β Zr transition phase and into base β-Zr material. The change of mechanical properties relates to oxygen concentration profiles obtained by wavelength dispersive spectroscopy (WDS) measurements.

Název v anglickém jazyce

Local mechanical properties testing of zirconium alloy nuclear fuel claddings

Popis výsledku anglicky

Zirconium alloys are being commonly used as a material of choice for nuclear fuel claddings in water cooled nuclear reactors for decades due to their good corrosion resistance and low neutron absorption. However, the increasing operation conditions of next generation nuclear reactors (Gen-IV) in terms of higher temperatures, pressures and higher neutron flux requires evaluation of further Zr cladding usability. The embrittlement of Zr claddings due to hydrogen pickup from reactor coolant is one of the issues for its potential use in Gen-IV reactors. Nanoindentation is an effective tool for analysis of the change of mechanical properties of hydrogen enriched Zr claddings from localised material volume. Zirconium alloy Zr-1Nb (E110) with experimentally induced hydrides was analysed by the means of nanoindentation. Zirconium hydrides were formed in the material after exposure in high temperature water autoclave. The nanoindentation measurements were performed as an array of 10x10 indents across areas with hydrides. Depth dependent mechanical properties measured by nanoindentation were compared between the material with no hydrogen content, low hydrogen content (127 ppm H) and high hydrogen content (397 ppm H). Another set of nanoindentation measurements were performed on Zircalloy-4 (Zr-1.5Sn-0.2Fe-0.1Cr) cladding tubes after high-temperature oxidation (950°C – 1425°C) in steam, simulating severe accident conditions in pressurised water reactors (PWRs). Linear nanoindentation profile measurements were conducted in radial direction of the cladding across ZrO2 layer, oxygen enriched α-Zr phase, α-β Zr transition phase and into base β-Zr material. The change of mechanical properties relates to oxygen concentration profiles obtained by wavelength dispersive spectroscopy (WDS) measurements.

Druh

O - Ostatní výsledky

CEP obor

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OECD FORD obor

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

Projekt

<a href="/cs/project/TK03020169" target="_blank" >TK03020169: Prostředky a metodiky pro kvalifikaci „Accident Tolerant“ pokrytí jaderného paliva</a><br>

Návaznosti

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

Rok uplatnění

2022

Kód důvěrnosti údajů

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