Hostname: page-component-76fb5796d-2lccl Total loading time: 0 Render date: 2024-04-25T21:08:33.915Z Has data issue: false hasContentIssue false
  • English
  • Français

Effect of Spent Fuel Burnup and Composition on Alteration of the U(Pu)O2 Matrix

Published online by Cambridge University Press:  21 March 2011

C. Jégou ,
S. Peuget ,
J.F. Lucchini ,
C. Corbel ,
V. Broudic  and
J.M. Bart
C. Jégou
Affiliation:
Commissariat à l'Énergie Atomique (CEA), Rhô Valley Research Center, DRRV/SCD, BP 171, 30207 Bagnols-sur-Cèze Cedex, France
S. Peuget
Affiliation:
Commissariat à l'Énergie Atomique (CEA), Rhô Valley Research Center, DRRV/SCD, BP 171, 30207 Bagnols-sur-Cèze Cedex, France
J.F. Lucchini
Affiliation:
Commissariat à l'Énergie Atomique (CEA), Rhô Valley Research Center, DRRV/SCD, BP 171, 30207 Bagnols-sur-Cèze Cedex, France
C. Corbel
Affiliation:
Commissariat à l'Énergie Atomique (CEA Saclay), DSM/DRECAM/SCM Laboratoire de radiolyse, 91191 Gif sur Yvette Cedex, France
V. Broudic
Affiliation:
Commissariat à l'Énergie Atomique (CEA), Rhô Valley Research Center, DRRV/SCD, BP 171, 30207 Bagnols-sur-Cèze Cedex, France
J.M. Bart
Affiliation:
Commissariat à l'Énergie Atomique (CEA), Rhô Valley Research Center, DRRV/SCD, BP 171, 30207 Bagnols-sur-Cèze Cedex, France
Get access

Abstract

For a potential performance assessment of direct disposal of spent fuel in a nuclear waste repository, the chemical reactions between the fuel and possible intruding water must be understood and the resulting radionuclide release must be quantified.

Leaching experiments were performed with five spent fuel samples from French power reactors (four UO2 fuel samples with burnup ratings of 22, 37, 47 and 60 GWd·tHM−1 and a MOX fuel sample irradiated to 47 GWd·tHM−1) to determine the release kinetics of the matrix containing most (over 95%) of the radionuclides. The experiments were carried out with granitic groundwater on previously leached sections of clad fuel rods in static mode, in an aerated medium at room temperature (25°C) in a hot cell.

After 1000 or 2000 days of leaching, the Sr/U congruence ratios for all the UO2 fuel samples ranged from 1 to 2; allowing for the experimental uncertainty, strontium can thus be considered as a satisfactory matrix alteration tracer. No significant burnup effect was observed on the alteration of the UO2 fuel matrix. The daily strontium release factor was approximately 1 à 10−7 d−1 for UO2 fuel, and five to six times higher for MOX fuel. Several alteration mechanisms (radiolysis, solubility, precipitation/clogging) are examined to account for the experimental findings.

Type
Research Article
Information
MRS Online Proceedings Library (OPL) , Volume 663: Symposium – Scientific Basis for Nuclear Waste management XXIV , 2000 , 399
Copyright
Copyright © Materials Research Society 2001

Access options

Get access to the full version of this content by using one of the access options below. (Log in options will check for institutional or personal access. Content may require purchase if you do not have access.)

References

REFERENCES

1. Poinssot, C., Jégou, C., Toulhoat, P., Piron, J.P., Gras, J.M., New Approach for the RN Source Terms from Spent Nuclear Fuel in Geological Disposal Conditions. These proceedings. MRS 2000, Sydney.Google Scholar
2. Jégou, C., Peuget, S., Bart, J.M., Gavazzi, A., Sambugaro, D., Leaching under dynamic conditions (Soxhlet Tests) and microstructural analysis of used UOX (62800MWd/t) and MOX (47500MWd/t) fuels. International meeting Atalante 2000. October 2000. Palais des Papes, Avignon, France. In prep.Google Scholar
3. , Hochanadel, J. Phys. Chem. 56, 587 (1952).Google Scholar
4. Forsyth, R.S., The SKB Spent Fuel Corrosion Program, SKB Technical Report 97-25, December 1997.Google Scholar
5. Casas, I., Gimenez, J., Marti, V., Torrero, M.E., Pablo, J. De, Kinetic Studies of Unirradiated UO2 Dissolution under Oxidizing Conditions in Batch and Flow Experiments, Radiochimica Acta, 66/67, 23–3, 1994.Google Scholar
6. Pavageau, J., In Synthesis on the Spent nuclear fuel evolution, Ed. Poinssot, C. (1999), to be published.Google Scholar
7. Forsyth, R.S., Werme, L.O, Spent Fuel Corrosion and Dissolution. Journal of Nuclear Materials 190, 1992 3–19.Google Scholar
8. Grambow, B., Werme, L.O., Forsyth, R.S., Bruno, J., Constraints by Experimental Data for Modeling of Radionuclide Release from Spent Fuel, Mat. Res. Soc. Symp. Proc. vol. 176, 1990.Google Scholar
9. Gray, W.J., Strachan, D. M, UO2 Matrix Dissolution Rates and Grain Boundary Inventories of Cs, Sr, and Tc in Spent Fuel, Mat. Res. Soc. Symp. Proc. vol. 212, 1991.Google Scholar