Numerical Modelling of Radionuclide Migration in the Context of a Near-Surface LILW Disposal Facility - 18618
- Neodyme, 86 bis rue Amelot, 75011 Paris 1 (France)
- Institut de Radioprotection et de Surete Nucleaire - IRSN, B.P.17, 92262 Fontenay-aux-Roses Cedex (France)
The reference solution for the management of low and intermediate level short-lived waste (LILW) in France consists in their emplacement in concrete vaults built at the near-surface. The corresponding disposal facility under operation by the French waste management organization (Andra), is located in North-eastern France and planned to be closed in 2060. The safety case developed by Andra has to show the evidence that, amongst other aspects, the radionuclide transfer through the confinement barriers will occur in such a way that the human and environmental exposure will be as low as reasonably achievable (ALARA) and in accordance with the dose constraints. For this purpose, modelling is used to integrate uncertainties about the characteristics of concrete components, waste and water table and their evolution in time and space. At last, modelling is finally used to investigate the radiological impact of specific scenarios in which the failure of safety barriers occurs. In the frame of its technical review of the safety assessment, the French Institute for Radiological Protection and Nuclear Safety (IRSN) also performs such kind of numerical modelling of radionuclide fate from the waste to the hydrogeological outlets, in order to check whether the assessment performed by Andra takes into account sufficiently conservative assumptions. The present work aims to evaluate the role of components of the vaults on radionuclide transport during the post-closure period over a few thousand years. For this purpose, IRSN carries out numerical calculations with MELODIE software, which simulates water flow and transport of species, in order to simulate radionuclide migration from waste packages emplaced in a concrete vault to the surrounding groundwater. The developed transport model focuses on 3-H radionuclide and integrates Andra's main assumptions (release model from the waste packages, chemical and hydraulic properties of the vault components, shrinkage and cracks in concrete components). An adaptive mesh method, based on an a posteriori error estimator, was used to model the cracks and the consideration of a strong heterogeneity in the hydraulic and transport parameters values in the different components of the model. This method shows an improvement of results of about 40 times, whereas the mesh size was increased by only 2%. (authors)
- Research Organization:
- WM Symposia, Inc., PO Box 27646, 85285-7646 Tempe, AZ (United States)
- OSTI ID:
- 22977867
- Report Number(s):
- INIS-US-20-WM-18618; TRN: US21V0495017912
- Resource Relation:
- Conference: WM2018: 44. Annual Waste Management Conference, Phoenix, AZ (United States), 18-22 Mar 2018; Other Information: Country of input: France; 4 refs.; Available online at: https://www.xcdsystem.com/wmsym/2018/index.html
- Country of Publication:
- United States
- Language:
- English
Similar Records
Modelling approach to LILW-SL repository safety evaluation for different waste packing options
Generic Argillite/Shale Disposal Reference Case
Related Subjects
ALARA
COMPUTER CODES
COMPUTERIZED SIMULATION
CONCRETES
FORECASTING
FRANCE
GROUND DISPOSAL
GROUND WATER
INTERMEDIATE-LEVEL RADIOACTIVE WASTES
LEAKS
LOW-LEVEL RADIOACTIVE WASTES
RADIATION PROTECTION
RADIOACTIVE WASTE DISPOSAL
RADIONUCLIDE MIGRATION
TRANSPORT THEORY