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Dissolution of spent nuclear fuel fragments at high alkaline conditions under H2 overpressure

Published online by Cambridge University Press:  20 February 2017

E. González-Robles ,
M. Herm ,
V. Montoya ,
N. Müller ,
B. Kienzler ,
R. Gens  and
V. Metz
E. González-Robles*
Affiliation:
Karlsruhe Institute of Technology, Institute for Nuclear Waste Disposal (KIT-INE), P.O. Box 3640, 76021 Karlsruhe, Germany
M. Herm
Affiliation:
Karlsruhe Institute of Technology, Institute for Nuclear Waste Disposal (KIT-INE), P.O. Box 3640, 76021 Karlsruhe, Germany
V. Montoya
Affiliation:
Karlsruhe Institute of Technology, Institute for Nuclear Waste Disposal (KIT-INE), P.O. Box 3640, 76021 Karlsruhe, Germany
N. Müller
Affiliation:
Karlsruhe Institute of Technology, Institute for Nuclear Waste Disposal (KIT-INE), P.O. Box 3640, 76021 Karlsruhe, Germany
B. Kienzler
Affiliation:
Karlsruhe Institute of Technology, Institute for Nuclear Waste Disposal (KIT-INE), P.O. Box 3640, 76021 Karlsruhe, Germany
R. Gens
Affiliation:
Organisme National des Déchets Radioactifs et des Matières Fissiles Enrichies / Nationale Instelling voor Radioactief Afval en Verrijkte Splijtstoffen (ONDRAF/NIRAS), Avenue des Arts 14, 1210 Brussels, Belgium
V. Metz
Affiliation:
Karlsruhe Institute of Technology, Institute for Nuclear Waste Disposal (KIT-INE), P.O. Box 3640, 76021 Karlsruhe, Germany
*
*(Email: [email protected])
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Abstract

The long-term behavior of the UO2 fuel matrix under conditions of the Belgian “Supercontainer design” was investigated by dissolution tests of high burn-up spent nuclear fuel (SNF) in high alkaline solution under 40 bar of (Ar + 8%H2) atmosphere. Four fragments of SNF, obtained from a pellet previously leached during two years, were exposed to young cement water with Ca (YCWCa) under 3.2 bar H2 partial pressure in four single/independent autoclave experiments for a period of 59, 182, 252 and 341 days, respectively. After a decrease of the concentration of dissolved 238U, which is associated with a reduction of U(VI) to U(IV), the concentration of 238U in solution is constant in the experiments running for 252 and 341 days. These observations indicate an inhibition of the matrix dissolution due to the presence of H2. A slight increase in the concentration of 90Sr and 137Cs in the aqueous solution indicates that there is still dissolution of the grain boundaries. These findings are similar to those reported for spent nuclear fuel corrosion in synthetic near neutral pH solutions.

Keywords

Corrosionwaste managementnuclear materials
Type
Articles
Information
MRS Advances , Volume 1 , Issue 62: Scientific Basis for Nuclear Waste Management XXXIX , 2016 , pp. 4163 - 4168
Copyright
Copyright © Materials Research Society 2017 

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References

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