CVR activities on the neutron irradiated fuel cladding materials

Kód výsledku v IS VaVaI

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Výsledek na webu

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DOI - Digital Object Identifier

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Jazyk výsledku

angličtina

Název v původním jazyce

CVR activities on the neutron irradiated fuel cladding materials

Popis výsledku v původním jazyce

This contribution describes the activities dedicated to the fuel cladding investigations performed in CVR. These activities include the studies of Zr-1Nb cladding tubes used in Czech NPPs as well as promising new accident tolerant fuel (ATF) cladding materials. Since 2007, CVR with its infrastructure built within the SUStainable ENergy Project started to provide the pre-commercial research for Czech nuclear power plants and successively participate in several international nuclear research projects. Among other projects, we work in a close cooperation with ÚJV Řež and UJP PRAHA on establishing and developing of new testing techniques for fuel cladding characterisation in the SUSEN hot cells. In order to support the safe and reliable operation of the nuclear fuels, experimental data must be acquired and used in both fuel performance monitoring and modelling. Different methods of mechanical testing, including tensile and creep studies supported by microstructural characterization (SEM, TEM), are applied to the zirconium alloys to predict their behaviour under various operating conditions after neutron irradiation. The studies focus on the effect of hydrogen in the fuel cladding, effects of the neutron irradiation and effects of strain aging on the tensile and creep behaviour, as well as a simulation of the fuel-cladding interaction and low- and high-cycle fatigue. Since the damage of the cladding by foreign objects is still an important operational issue, CVR is working on developing the techniques for fretting experiments on fuel cladding materials with protective coatings. The development of the procedure includes the simulation of the friction of a foreign object against the coated cladding tube in dry and aqueous environment. Furthermore, the depth of the resulting groove is measured using microscopy or profilometer. The aim of the tests is to show the differences between the reference samples without coating and samples of the same alloy with different types of coatings and to demonstrate the positive effect of the coatings on the cladding resistance. As part of the development of the procedure, an autoclave is also being developed to ensure elevated temperatures and chemical regime during testing and to be able to test irradiated samples at a later stage. Presented at HOTLAB 2021 conference.

Název v anglickém jazyce

CVR activities on the neutron irradiated fuel cladding materials

Popis výsledku anglicky

This contribution describes the activities dedicated to the fuel cladding investigations performed in CVR. These activities include the studies of Zr-1Nb cladding tubes used in Czech NPPs as well as promising new accident tolerant fuel (ATF) cladding materials. Since 2007, CVR with its infrastructure built within the SUStainable ENergy Project started to provide the pre-commercial research for Czech nuclear power plants and successively participate in several international nuclear research projects. Among other projects, we work in a close cooperation with ÚJV Řež and UJP PRAHA on establishing and developing of new testing techniques for fuel cladding characterisation in the SUSEN hot cells. In order to support the safe and reliable operation of the nuclear fuels, experimental data must be acquired and used in both fuel performance monitoring and modelling. Different methods of mechanical testing, including tensile and creep studies supported by microstructural characterization (SEM, TEM), are applied to the zirconium alloys to predict their behaviour under various operating conditions after neutron irradiation. The studies focus on the effect of hydrogen in the fuel cladding, effects of the neutron irradiation and effects of strain aging on the tensile and creep behaviour, as well as a simulation of the fuel-cladding interaction and low- and high-cycle fatigue. Since the damage of the cladding by foreign objects is still an important operational issue, CVR is working on developing the techniques for fretting experiments on fuel cladding materials with protective coatings. The development of the procedure includes the simulation of the friction of a foreign object against the coated cladding tube in dry and aqueous environment. Furthermore, the depth of the resulting groove is measured using microscopy or profilometer. The aim of the tests is to show the differences between the reference samples without coating and samples of the same alloy with different types of coatings and to demonstrate the positive effect of the coatings on the cladding resistance. As part of the development of the procedure, an autoclave is also being developed to ensure elevated temperatures and chemical regime during testing and to be able to test irradiated samples at a later stage. Presented at HOTLAB 2021 conference.

Druh

O - Ostatní výsledky

CEP obor

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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

2021

Kód důvěrnosti údajů

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