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Modelling Approach to LILW-SL Repository Safety Evaluation for Different Waste Packing Options

[+] Author Affiliations
Janez Perko, Dirk Mallants, Geert Volckaert

SCK•CEN, Belgium

George Towler, Mike Egan

Quintessa, Henley-on-Thames, Oxfordshire, England, UK

Sandi Viršek, Bojan Hertl

ARAO, Slovenia

Paper No. ICEM2007-7069, pp. 1189-1196; 8 pages
  • The 11th International Conference on Environmental Remediation and Radioactive Waste Management
  • 11th International Conference on Environmental Remediation and Radioactive Waste Management, Parts A and B
  • Bruges, Belgium, September 2–6, 2007
  • Conference Sponsors: Nuclear Division and Environmental Engineering Division
  • ISBN: 978-0-7918-4339-0 | eISBN: 0-7918-3818-8
  • Copyright © 2007 by ASME


The key objective of the work described here was to support the identification of a preferred disposal concept and packaging option for low and short-lived intermediate level waste (LILW-SL). The emphasis of the assessment, conducted on behalf of the Slovenian radioactive waste management agency (ARAO), was the consideration of several waste treatment and packaging options in an attempt to identify optimised containment characteristics that would result in safe disposal, taking into account the cost-benefit of alternative safety measures. Waste streams for which alternative treatment and packaging solutions were developed and evaluated include decommissioning waste and NPP operational wastes, including drums with unconditioned ion exchange resins in overpacked tube type containers (TTCs). For decommissioning wastes, the disposal options under consideration were either direct disposal of loose pieces grouted into a vault or use of high integrity containers (HIC). In relation to operational wastes, three main options were foreseen. The first is overpacking of resin containing TTCs grouted into high integrity containers, the second option is complete treatment with hydration, neutralization, and cementation of the dry resins into drums grouted into high integrity containers and the third is direct disposal of TTCs into high integrity containers without additional treatment. The long-term safety of radioactive waste repositories is usually demonstrated with the support of a safety assessment. This normally includes modelling of radionuclide release from a multi-barrier near-surface or deep repository to the geosphere and biosphere. For the current work, performance assessment models were developed for each combination of siting option, repository design and waste packaging option. Modelling of releases from the engineered containment system (the ‘near-field’) was undertaken using the AMBER code [1]. Detailed unsaturated water flow modelling was undertaken using the HYDRUS code [2], where the degree of engineered barrier degradation with time is accounted for in each packaging option. Water fluxes relating to each degradation level were then incorporated into the AMBER models for further radionuclide transport calculations appropriate to each packing solution. The approach proved to be highly flexible, transparent and effective in terms of calculation time. Results demonstrate that all waste streams could be accepted at the preferred site with the surface repository option, under the condition that all decommissioning waste would be grouted into high integrity containers. The use of high integrity containers is also recommended for all other waste streams. Results from the detailed analysis further showed that in-drum-dried ion exchange resins in TTCs would be acceptable when grouted into high integrity containers, thereby avoiding the need for complicated processing and repackaging.

Copyright © 2007 by ASME



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