Hostname: page-component-848d4c4894-x24gv Total loading time: 0 Render date: 2024-05-06T03:18:07.528Z Has data issue: false hasContentIssue false
  • English
  • Français

Static Dissolution of α-doped UO2 in Boom Clay Conditions: Preliminary Results

Published online by Cambridge University Press:  01 February 2011

C. Cachoir  and
K. Lemmens
C. Cachoir
Affiliation:
Waste and Disposal Department, SCK•CEN, Boeretang 200, B-2400 Mol, Belgium.
K. Lemmens
Affiliation:
Waste and Disposal Department, SCK•CEN, Boeretang 200, B-2400 Mol, Belgium.
Get access

Abstract

To support the performance assessment (PA) calculations for the possible final disposal of spent fuel in the Boom Clay, leach experiments are performed with α-doped UO2 in clay media, simulating various near field ages. The experiments allow to measure the dissolution rate of the UO2 and to determine the assumed relationship between dissolution rate and α-activity. Tests are performed at six different α-activities, simulating various fuel ages, at 25–30°C, for durations ranging from 90 to 720 days, in a glove box with Ar/0.4%CO2 atmosphere. The solutions and solids are analyzed for U isotopes and 238Pu by radiochemical measurement and by ICP-MS. The dissolution rates of the α-doped UO2 are presented for different durations. The resulting corrosion rate is around 300 μgU.m−2.d−1. This is up to 100 times higher than found for similar conditions in the literature. In the presence of clay, there appears to be no correlation between the α-activity and the corrosion rate of the α-doped UO2.

Type
Research Article
Information
MRS Online Proceedings Library (OPL) , Volume 807: Symposium Scientific Basis for Nuclear Waste Management XXVII , 2003 , 59
Copyright
Copyright © Materials Research Society 2004

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. SAFIR 2. Safety Assessment and Feasibility Interim Report 2, NIRAS/ONDRAF publications 2001–06 E (2001).Google Scholar
2. Source term for performance assessment of spent fuel as a waste form, Final report Contract FI4W-CT-95–0004, (2000) EUR 19140.Google Scholar
3. Marivoet, J., Volckaert, G., Sneyers, A., Wibin, J., First performance assessment of the disposal of spent fuel in a clay layer, Final report for contract FI2W-CT90–0016 with the EC and contract CCHO-89–094 with NIRAS/ONDRAF, (1996), EUR 16752.Google Scholar
4. SCK·CEN's contribution to Topical report 1 of the SPA project (1998).Google Scholar
5. Rondinella, V., Matzke, Hj, Cobos, J., Wiss, T., α-radiolysis and α-radiation damage effects on UO2 dissolution under spent fuel storage conditions. Mat. Res. Soc. Symp. Proc. Vol 556, (1999), 447–454.Google Scholar
6. Oversby, V.. Uranium dioxide, SIMFUEL, and spent fuel dissolution rates- a review of published data. SKB report TR-99–22, October 1999.Google Scholar
7. Lajudie, A., Raynal, J., Petit, J.C. and Toulhoat, P.. Clay-based materials for engineered barriers: a review. Mat. Res. Symp. Proc. Vol. 353, 1995, pp 221–230.Google Scholar
8. Lemmens, K., Van Iseghem, P., Wang, ı.. The leaching of Pu, Am, Np and Tc from high-level waste glasses in clay media. Mater. Res. Soc. Symp. Proc. 294 (1993) 147.Google Scholar
9. Sillen, X., Marivoet, J.. Spent Fuel performance assessment for a hypothetical repository in the Boom Clay at the Mol site (Belgium). SCK•CEN report BLG-877 (2002).Google Scholar
10. Vieno, T., Nordman, H.. TILA-96: Interim report on safety assessment of spent fuel disposal. POSIVA, Helsinki, Report POSIVA-96–17 (1996).Google Scholar
11. Bonnal, M., Charles, B., Broudic, V. and Jégou, C.. Alpha and gamma radiolysis effects on UO2 alteration in water. Spent fuel Workshop 23–25 September 2002, Avignon-France.Google Scholar
12. Le Lous, K.. Etude de la lixiviation du combustible usé en présence de matériaux d'environnement. Thèse de doctorat de l'Université Paris Sud (France), 1997.Google Scholar
13. Kim, S. S., Chun, K.S. and Kang, C.H.. The release of radionuclides from spent fuel in bentonite block. Spent fuel Workshop 23–25 September 2002, Avignon-France.Google Scholar