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Radionuclides Release Model for Performance Assessment Studies of Spent Nuclear Fuel in Geological Disposal

Published online by Cambridge University Press:  17 March 2011

Christophe Poinssot ,
Patrick Lovera  and
Cécile Ferry
Christophe Poinssot
Affiliation:
CEA Saclay, Nuclear Energy Division, Department of Physics and Chemistry, Service for the Studies of Radionuclides Behavior, BP.11, F-91191 Gif-sur-Yvette Cedex, France;, christophe.poinssot@cea.fr
Patrick Lovera
Affiliation:
CEA Saclay, Nuclear Energy Division, Department of Physics and Chemistry, Service for the Studies of Radionuclides Behavior, BP.11, F-91191 Gif-sur-Yvette Cedex, France
Cécile Ferry
Affiliation:
CEA Saclay, Nuclear Energy Division, Department of Physics and Chemistry, Service for the Studies of Radionuclides Behavior, BP.11, F-91191 Gif-sur-Yvette Cedex, France
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Abstract

In the framework of the research conducted on the long term evolution of spent nuclear fuel in geological disposal conditions, a source term model has been developed to evaluate the instantaneous release of RN (Instant Release Fraction IRF) and the delayed release of the RN which are embedded within the matrix. This model takes into account all the scientific results currently available in the literature except the hydrogen effect. IRF was assessed by considering the evolution with time of the RN inventories located within the fuel microstructure to which no confinement properties can be allocated on the long term (rim, gap, grain boundaries). It allows to propose some reference bounding values for the IRF as a function of time of canister breaching and burnup. The matrix radiolytic dissolution was modeled by a simple kinetic model neglecting the radiolytic species recombination and the influence of aqueous ligands and radiolytic oxidants were supposed to completely react with the fuel surface. Spent fuel performance was therefore demonstrated to deeply depend on the reactive surface area.

Type
Research Article
Information
MRS Online Proceedings Library (OPL) , Volume 824: Symposium CC – Scientific Basis for Nuclear Waste Management XXVIII , 2004 , CC2.2
Copyright
Copyright © Materials Research Society 2004

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