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The Source Term for the Release of Nuclides From a Radioactive Waste Repository – 1. Vitrified waste in granite

Published online by Cambridge University Press:  25 February 2011

F.T. Ewart ,
J.B. Morris ,
J. Severn ,
B.M. Sharpe  and
H.P. Thomason
F.T. Ewart
Affiliation:
Chemical Technology Division, Atomic Energy Research Establishment, Harwell, Oxford, OXll ORA, U.K.
J.B. Morris
Affiliation:
Chemical Technology Division, Atomic Energy Research Establishment, Harwell, Oxford, OXll ORA, U.K.
J. Severn
Affiliation:
Chemical Technology Division, Atomic Energy Research Establishment, Harwell, Oxford, OXll ORA, U.K.
B.M. Sharpe
Affiliation:
Chemical Technology Division, Atomic Energy Research Establishment, Harwell, Oxford, OXll ORA, U.K.
H.P. Thomason
Affiliation:
Chemical Technology Division, Atomic Energy Research Establishment, Harwell, Oxford, OXll ORA, U.K.
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Abstract

Samples of simulated vitrified radioactive waste have been prepared containing, separately, neptunium, plutonium and technetium. These glasses have been leached in static granite groundwater for 150 days. The resulting solutions have been characterised with respect to both the ionic and colloidal species: these solutions have then been contacted with concrete and granite to produce initial measurements of the source term from a typical repository. The results show that (a) the colloidal species arise from detached particles of the so-called gel layer on the surface of the glass, (b) e solution activities were 590 Bq.ml-1 in the case of 99Tc and 20 Bq.ml-1 for 237Np, and (c) the sorption coefficients wIre 1.4 and > 300 ml.g for concrete and 0.4 and 8.3 ml.g-1 for granite respectively. The technetium was only weakly sorbed and the neptunium strongly sorbed in each case.

Type
Research Article
Information
MRS Online Proceedings Library (OPL) , Volume 26: Symposium D – Scientific Basis for Nuclear Waste Management VII , 1983 , 171
Copyright
Copyright © Materials Research Society 1984

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References

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