Modelling experimental results on radiolytic processes at the spent fuel water interface. II. Radionuclides release
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Modelling experimental results on radiolytic processes at the spent fuel water interface. II. Radionuclides release. E. Cera1, M. Grivé1, J. Bruno1, T.E. Eriksen2 1 Enviros Spain SL, Pg. de Rubà 29-31, 08197 Valldoreix (Spain) 2 Dept. Nuclear Chemistry, KTH, 100 44 Stockholm (Sweden) ABSTRACT Experimental and modelling efforts in the last decade in the frame of nuclear waste management field have been focused on studying the role of the UO2 surfaces in poising the redox state of solid/water systems as well as the radionuclides release behaviour. For this purpose, an experimental programme was developed consisting on dissolution experiments with PWR spent fuel fragments in an anoxic environment and by using different solution compositions. Some of the collected data has been previously published [1], specifically those data concerning radiolysis products and dissolution of the matrix. The results and the modelling tasks indicated an overall balance of the generated radiolytic species and that uranium dissolution was controlled by the oxidation of the spent fuel matrix in 10mM bicarbonate solutions while in the tests carried out at lower or without carbonate concentrations uranium in the aqueous phase was governed by the precipitation of schoepite. This paper is the continuation of a series accounting for the data and modelling work related to investigating the release behaviour of minor radionuclides from the spent fuel. Uranium concentrations as a function of time showed an initial increase until reaching a steady state, indicating a matrix dissolution control. The same behaviour is observed for neptunium, caesium, strontium, technetium and molybdenum indicating a congruent release of these elements with the major component of the fuel matrix. On the other hand, no clear tendency is observed for plutonium data where additional solubility limiting mechanisms may apply. Kinetic modelling of the trace elements: caesium, strontium, technetium and molybdenum is based on the congruent release of these elements with the major component of the fuel matrix. Rate constants have been determined. Kinetic modelling of neptunium data took also into account the subsequent precipitation as Np(IV) hydroxide. Finally, measured Pu concentrations may be explained by the precipitation of Pu(IV) and/or Pu(III) solid phases.
INTRODUCTION The spent fuel matrix is the first barrier within the repository design given the high stability of this material in an anoxic media. The confinement of radionuclides within the matrix is guaranteed if the oxidation state of the UO2 matrix does not exceed the upper limit of stability of the cubic structure, UO2.33, which corresponds to a nominal stoichiometry of U3O7 [2,3], consequently its stability will depend on the oxidant species able to oxidise the UO2 matrix until an oxidation state above this upper limit. Experimental and modelling efforts in the last decade in the frame of nuclear waste management field have been focused on studying the role of the UO2
surfaces in poising the redox state of so
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