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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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Metz, Volker
Karlsruhe Institute of Technology
in Cooperation with on an Cooperation-Score of 37%
Topics
Publications (7/7 displayed)
- 2021Anaerobic corrosion of carbon steel in compacted bentonite exposed to natural opalinus clay porewater: bentonite alternation study
- 2021Reactive transport modelling of the long-term interaction between carbon steel and MX-80 bentonite at 25 °Ccitations
- 2021Cement as a component of the multi-barrier system: radionuclide retention in cementitious environmentscitations
- 2021Anaerobic corrosion of carbon steel in compacted bentonite exposed to natural Opalinus clay porewater: Bentonite alteration studycitations
- 2021Twenty years of research on spent nuclear fuel in the context of final disposal at KIT-INE in multinational collaborative projects
- 2018Overview of ¹⁴C release from irradiated zircaloys in geological disposal conditionscitations
- 2012Corrosion of spent nuclear fuel segment in presence of Fe(II)/Fe(III) oxide
Places of action
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article
Overview of ¹⁴C release from irradiated zircaloys in geological disposal conditions
Abstract
Carbon-14 (radiocarbon, 14C) is a long-lived radionuclide (5730 yr) of interest regarding the safety for the management of intermediate level wastes (ILW). The present study gives an overview of the release of 14C from irradiated Zircaloy cladding in alkaline media. 14C is found either in the alloy part of Zircaloy cladding due to the neutron activation of 14N impurities by 14N(n,p)14C reaction, or in the oxide layer (ZrO2) formed at the metal surface by the neutron activation of 17O from UO2 or (U-Pu)O2 fuel and water from the primary circuit in the reactor by 17O(n,α)14C reaction. Various irradiated and unirradiated Zircaloys have been studied. The total 14C inventory has been determined both experimentally and by calculations. The results seem to be in good agreement. Leaching experiments were conducted in alkaline media for several time durations. 14C was mainly released as carboxylic acids. Further, corrosion measurements were performed by using both hydrogen measurements and electrochemical measurements. The corrosion rate (CR) ranges from a few nm/yr to 100 nm/yr depending on the surface conditions and the method used for measurement. From a safety assessment point of view, the instant release fraction (IRF) was determined on irradiated Zircaloy-2. The results showed that the 14C inventory in the oxide was significantly below the 20% commonly used in safety case assessments.