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An Overview of Mechanisms of the Degradation of Promising ATF Cladding Materials During Oxidation at High Temperatures

High temperature corrosion of materials, 2024, Vol.101 (4), p.621-647

2024

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The SPIZWURZ project – Experimental investigations and modeling of the behavior of hydrogen in zirconium alloys under long-term dry storage conditions

Nuclear Engineering and Technology, 2024, 56(3), , pp.824-831

2024

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Experimental and modelling results of the QUENCH-20 experiment with BWR test bundle

Nuclear engineering and design, 2023-08, Vol.410, p.112391, Article 112391

2023

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Experimental and modelling results of the QUENCH-18 bundle experiment on air ingress, cladding melting and aerosol release

Nuclear engineering and design, 2021-08, Vol.379, p.111267, Article 111267

2021

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Post-quench ductility limits of coated ATF with various zirconium-based alloys and coating designs

Journal of nuclear materials, 2024-04, Vol.591, p.154915, Article 154915

2024

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Preliminary validation of ASTEC V2.2.b with the QUENCH-20 BWR bundle experiment

Nuclear engineering and design, 2020-12, Vol.370, p.110931, Article 110931

2020

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Microstructure and mechanical properties of Zircaloy-4 cladding hydrogenated at temperatures typical for loss-of-coolant accident (LOCA) conditions

Nuclear engineering and design, 2015-03, Vol.283, p.33-39

2015

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Analysis of iron-chromium-aluminum samples exposed to accident conditions followed by quench in the QUENCH-19 experiment

Journal of nuclear materials, 2023-04, Vol.580 (1)

2023

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Analysis of iron-chromium-aluminum samples exposed to accident conditions followed by quench in the QUENCH-19 experiment

Journal of nuclear materials, 2023-07, Vol.580 (C), p.154433, Article 154433

2023

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Hafnium oxidation at high temperature in steam

Journal of nuclear materials, 2021-07, Vol.550, p.152901-14, Article 152901

2021

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Hydride precipitation, fracture and plasticity mechanisms in pure zirconium and Zircaloy-4 at temperatures typical for the postulated loss-of-coolant accident

Nuclear engineering and design, 2016-05, Vol.301, p.366-377

2016

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The SPIZWURZ project - Experimental investigations and modeling of the behavior of hydrogen in zirconium alloys under long-term dry storage conditions

Nuclear engineering and technology, 2024, Vol.56 (3), p.824-831

2024

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The measurement of Ag/In/Cd release under air-ingress conditions in the QUENCH-18 bundle test

Journal of nuclear materials, 2019-04, Vol.517, p.315-327

2019

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Effect of steam and oxygen starvation on severe accident progression with air ingress

Nuclear engineering and design, 2022-09, Vol.396, p.111884, Article 111884

2022

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Application of Thermal Hydraulic and Severe Accident Code SOCRAT/V3 to Bottom Water Reflood Experiment QUENCH-LOCA-1

Journal of nuclear engineering and radiation science, 2016-04, Vol.2 (2)

2016

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Influence of oxide layer morphology on hydrogen concentration in tin and niobium containing zirconium alloys after high temperature steam oxidation

Journal of nuclear materials, 2009-03, Vol.385 (2), p.339-345

2009

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Secondary hydriding during LOCA – Results from the QUENCH-L0 test

Journal of nuclear materials, 2012, Vol.420 (1), p.575-582

2012

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On the EU-Japan roadmap for experimental research on corium behavior

Annals of nuclear energy, 2019-02, Vol.124, p.541-547

2019

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Application of neutron radiography to study material processes during hypothetical severe accidents in nuclear reactors

Journal of materials science, 2012-09, Vol.47 (18), p.6505-6512

2012

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Validation of ASTECV2.1 based on the QUENCH-08 experiment

Nuclear engineering and design, 2017-04, Vol.314, p.29-43

2017

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Land contamination activity data interpretation from Fukushima Daiichi accident

Nuclear engineering and design, 2016-04, Vol.300, p.28-33

2016

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SARNET2 benchmark on air ingress experiments QUENCH-10, -16

Annals of nuclear energy, 2014-12, Vol.74 (C), p.12-23

2014

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ICONE23-1724 HYDRIDES AND FRACTURE OF PURE ZIRCONIUM AND ZIRCALOY - 4 HYDROGENATED AT TEMPERATURES TYPICAL FOR LOSS-OF-COOLANT ACCIDENT CONDITIONS

The Proceedings of the International Conference on Nuclear Engineering (ICONE), 2015, Vol.2015.23, p.ICONE23-1-ICONE23-1

2015

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Experiments on air ingress during severe accidents in LWRS

Nuclear engineering and design, 2006-08, Vol.236 (14), p.1709-1719

2006

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Aerosol behavior during SIC control rod failure in QUENCH-13 test

Journal of nuclear materials, 2010-02, Vol.397 (1), p.92-100

2010

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An Overview of Mechanisms of the Degradation of Promising ATF Cladding Materials During Oxidation at High Temperatures

The SPIZWURZ project – Experimental investigations and modeling of the behavior of hydrogen in zirconium alloys under long-term dry storage conditions

Experimental and modelling results of the QUENCH-20 experiment with BWR test bundle

Experimental and modelling results of the QUENCH-18 bundle experiment on air ingress, cladding melting and aerosol release

Post-quench ductility limits of coated ATF with various zirconium-based alloys and coating designs

Preliminary validation of ASTEC V2.2.b with the QUENCH-20 BWR bundle experiment

Microstructure and mechanical properties of Zircaloy-4 cladding hydrogenated at temperatures typical for loss-of-coolant accident (LOCA) conditions

Analysis of iron-chromium-aluminum samples exposed to accident conditions followed by quench in the QUENCH-19 experiment

Analysis of iron-chromium-aluminum samples exposed to accident conditions followed by quench in the QUENCH-19 experiment

Hafnium oxidation at high temperature in steam

Hydride precipitation, fracture and plasticity mechanisms in pure zirconium and Zircaloy-4 at temperatures typical for the postulated loss-of-coolant accident

The SPIZWURZ project - Experimental investigations and modeling of the behavior of hydrogen in zirconium alloys under long-term dry storage conditions

The measurement of Ag/In/Cd release under air-ingress conditions in the QUENCH-18 bundle test

Effect of steam and oxygen starvation on severe accident progression with air ingress

Application of Thermal Hydraulic and Severe Accident Code SOCRAT/V3 to Bottom Water Reflood Experiment QUENCH-LOCA-1

Influence of oxide layer morphology on hydrogen concentration in tin and niobium containing zirconium alloys after high temperature steam oxidation

Secondary hydriding during LOCA – Results from the QUENCH-L0 test

On the EU-Japan roadmap for experimental research on corium behavior

Application of neutron radiography to study material processes during hypothetical severe accidents in nuclear reactors

Validation of ASTECV2.1 based on the QUENCH-08 experiment

Land contamination activity data interpretation from Fukushima Daiichi accident

SARNET2 benchmark on air ingress experiments QUENCH-10, -16

ICONE23-1724 HYDRIDES AND FRACTURE OF PURE ZIRCONIUM AND ZIRCALOY - 4 HYDROGENATED AT TEMPERATURES TYPICAL FOR LOSS-OF-COOLANT ACCIDENT CONDITIONS

Experiments on air ingress during severe accidents in LWRS

Aerosol behavior during SIC control rod failure in QUENCH-13 test

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