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| Naji, M. |
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| Motta, Antonella |
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| Aletan, Dirar |
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| Mohamed, Tarek |
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| Ertürk, Emre |
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| Taccardi, Nicola |
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| Kononenko, Denys |
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| Petrov, R. H. | Madrid |
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| Alshaaer, Mazen | Brussels |
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| Bih, L. |
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| Casati, R. |
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| Muller, Hermance |
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| Kočí, Jan | Prague |
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| Šuljagić, Marija |
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| Kalteremidou, Kalliopi-Artemi | Brussels |
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| Azam, Siraj |
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| Ospanova, Alyiya |
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| Blanpain, Bart |
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| Ali, M. A. |
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| Popa, V. |
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| Rančić, M. |
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| Ollier, Nadège |
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| Azevedo, Nuno Monteiro |
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| Landes, Michael |
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| Rignanese, Gian-Marco |
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Steinbrück, Martin
in Cooperation with on an Cooperation-Score of 37%
Topics
Publications (35/35 displayed)
- 2024Decoding the oxidation mechanism of Zircaloy-4 via in situ synchrotron X-ray diffraction and computational elucidation
- 2024Phase formation, structure and properties of quaternary MAX phase thin films in the Cr-V-C-Al system: A combinatorial study
- 2023Synthesis of V2AlC thin films by thermal annealing of nanoscale elemental multilayered precursors : Incorporation of layered Ar bubbles and impact on microstructure formationcitations
- 2023Synthesis of V$_{2}$AlC thin films by thermal annealing of nanoscale elemental multilayered precursors: Incorporation of layered Ar bubbles and impact on microstructure formation
- 2023Nitriding model for zirconium based fuel cladding in severe accident codescitations
- 2023Analysis of iron-chromium-aluminum samples exposed to accident conditions followed by quench in the QUENCH-19 experiment
- 2022Phase formation and thermal stability of quaternary MAX phase thin films in the Cr-V-C-Al system: an experimental combinatorial study
- 2022The Effect of Annealing Temperature on the Microstructure and Properties of Cr–C–Al Coatings on Zircaloy-4 for Accident-Tolerant Fuel (ATF) Applicationscitations
- 2022Oxidation of silicon carbide composites for nuclear applications at very high temperatures in steamcitations
- 2022Results of metallographic analysis of the QUENCH-20 bundle with B4C absorber
- 2022Results of metallographic analysis of the QUENCH-20 bundle with B₄C absorber
- 2021Development of Cr-C-Al based coatings for enhanced accident tolerant fuel (ATF) zirconium-based alloy cladding
- 2021High-temperature oxidation and hydrothermal corrosion of textured Cr$_{2}$AlC-based coatings on zirconium alloy fuel cladding
- 2020High-Temperature Oxidation of Chrome-Nickel Alloycitations
- 2020Investigation of corrosion and high temperature oxidation of promising ATF cladding materials in the framework of the Il trovatore project
- 2018High-temperature interaction of oxygen-preloaded Zr1Nb alloy with nitrogencitations
- 2018H2 PERMEATION BEHAVIOR OF Cr2AlC AND Ti2AlC MAX PHASE COATED ZIRCALOY-4 BY NEUTRON RADIOGRAPHYcitations
- 2018H2 PERMEATION BEHAVIOR OF Cr2AlC AND Ti2AlC MAX PHASE COATED ZIRCALOY-4 BY NEUTRON RADIOGRAPHYcitations
- 2018Magnetron-sputtered Al-containing MAX phase carbide thin films and their application as oxidation-resistant coatings
- 2017High-temperature interaction of oxygen-preloaded Zr1Nb alloy with nitrogen
- 2017Nanocrystalline diamond protects Zr cladding surface against oxygen and hydrogen uptake : Nuclear fuel durability enhancementcitations
- 2017Lessons learned from the QUENCH-LOCA experiments
- 2017Update of the QUENCH program
- 2017UK Research in LOCA-Related Activities and Co-operation with the Karlsruhe Research Centre - A Historical Perspective
- 2016Oxidation and hydrogen uptake during high-temperature reaction of zirconium alloys in steam-nitrogen mixtures
- 2016High-temperature oxidation of SiC-Ta-SiC sandwich cladding tubes in GFR atmosphere
- 2013Results of the QUENCH-16 Bundle Experiment on Air Ingress (KIT Scientific Reports ; 7634)
- 2012Oxidation of zirconium alloys in mixed atmospheres containing nitrogen
- 2012Selected aspects of materials behavior during severe nuclear accidents in nuclear reactors
- 2012High-temperature oxidation and mutual interactions of materials during severe nuclear accidents
- 2012Separate effects experiments in the framework of the QUENCH program at KIT
- 2012Materials behavior during the early phase of a severe nuclear accident
- 2011Results of Severe Fuel Damage Experiment QUENCH-15 with ZIRLO cladding tubes. (KIT Scientific Reports ; 7576)
- 2010Separate-effects tests on the investigation of high-temperature oxidation behavior and mechanical properties of Zircaloy-2 to be used in the SFP PWR tests : Report prepared in the framework of the OECD/NEA SFP Project
- 2007Prototypical experiments on air oxidation of zircaloy-4 at high temperatures
Places of action
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document
High-temperature oxidation and mutual interactions of materials during severe nuclear accidents
Abstract
Karlsruhe Institute of Technology, Institute for Applied Materials, Germany ABSTRACT (Invited Talk) During a nuclear accident with a loss of coolant, the reactor core steadily heats up due to the release of decay heat and reduced heat transfer to the remaining steam. The temperature rise extends up to the point where stability limits of some materials in the core structure are reached and complex chemical reactions are involved. Oxidation of zirconium alloy cladding material becomes significant from temperatures of about 1000°C, causing mechanical degradation and a loss-of-barrier (against release of fission products) effect. Furthermore, this reaction is strongly exothermal, i.e. connected with release of heat comparable to and exceeding the residual nuclear power; and it is the main source of hydrogen during a nuclear accident jeopardizing the containment and reactor building (as seen during the Fukushima Daiichi accidents) and may be absorbed by metallic zirconium. Nitrogen is used for inertization of boiling water reactor (BWR) containments and for pressurization of emergency cooling water systems and comes into play during air ingress scenarios. It strongly affects the oxidation kinetics by the formation of zirconium nitride and its re-oxidation. Due to the significantly different densities of ZrN and ZrO2, porous, non-protective oxide layers are formed over a wide temperature range. Depending on temperature, the oxidation of Zry in steam-nitrogen mixtures may be faster than the oxidation in steam by one order of magnitude. The various core component materials are chemically unstable with respect to each other and eutectic interactions occur which lead to the formation of liquid phases in LWR fuel rod bundles at temperatures of approx. 1200°C already, i.e. significantly below the melting temperatures of the materials involved. Initial degradation occurs in the control rods with Ag-In-Cd alloy and boron carbide absorber materials. Whereas the low-temperature Ag-In-Cd alloy (used in most PWRs; melting temperature about 800°C) does not interact chemically with the enclosing stainless steel cladding, very rapid eutectic interactions between B4C (used in BWRs, recent PWRs, and VVERs; melting temperature 2450°C) and stainless steel as well as between stainless steel and zircaloy take place at about 1250°C. Failure of the Ag-In-Cd control rods is caused by high Cd vapor pressure and/or eutectic interaction between the surrounding steal and Zircaloy tubes. The resulting absorber melt may attack adjacent fuel rods and is an additional source of hydrogen and heat due to its rapid oxidation. The paper discusses the materials interactions in the early phase of a severe nuclear accident and presents highlights of the corresponding research at KIT, including large-scale bundle experiments and separate-effects tests on the laboratory scale.