Comparative Analysis of Results From Deterministic Calculations of the Release of Radioactivity From Geological Disposal
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Comparative Analysis of Results From Deterministic Calculations of the Release of Radioactivity From Geological Disposal of Spent Fuel in Crystalline Rocks Slavka Prvakova, Pavel V. Amosov1 and Karl-Fredrik Nilsson EC-JRC, Institute for Energy, P.O. Box 2, 1755 ZGPetten, Netherlands. 1 Mining Institute, MI KSC, Fersman Street, 24, Apatity, Murmansk Region, 184200, Russia. ABSTRACT This paper presents results from deterministic calculations of radionuclide migration in a deep geological environment. The concept assumes single canister with spent nuclear fuel situated in bentonite (near-field) and surrounded by crystalline host rock (far-field). The results are presented in the form of release rates from the near-field and far-field, which contains also a part of advective release from the system of two single fractures. Additionally, radiological risk is evaluated in the form of dose rates for a water drinking scenario for an exposed population group. The analyses have been done by using the methodologies and computer codes used at the Institute for Energy EC-JRC in the Netherlands and the Mining Institute KSC RAS in Russia. INTRODUCTION Development of a safety case for disposal of radioactive waste involves consideration of the evolution of the waste and engineered barrier systems, and the interactions between these complex systems, which are also evolving. The presented paper compares results from deterministic calculations of the radionuclide releases from the hypothetical geological repository situated in crystalline host rock. Radionuclide migration in the deep geological environment was simulated as a one-dimensional model for the timescale of one million years over which the repository system has to provide well-functioning barriers against radionuclide transport. There are future plans to carry out uncertainty analyses of the stochastically varied input parameters carried out by probabilistic techniques, taking into account simultaneous variation of all input data. CONCEPTUAL MODEL The conceptual model (see figure 1) assumes a single canister with spent nuclear fuel placed in bentonite which is surrounded by crystalline host rock (granite). The repository is expected to be situated below the water table and so saturated conditions are assumed to prevail for the whole time of disposal. The source term consists of the radionuclides C-14, Cl-36, Ni-59, Se-79, Nb-94, Tc-99, I-129, Cs-135, Pu-239 and Am-243. The canister is assumed to fail completely 1000 years after disposal. No sorption is modelled in the region of the source term. Within both barriers, bentonite and part of the host rock, transport is assumed to be purely diffusive. Radionuclides can sorb onto these materials according to the material specific
distribution coefficients and may be subject to solubility constraints. Few of them, namely C-14, Cl-36 and I-129, can be subject to anion exclusion processes which reduces their effective porosity. The input data regarding the half-lives, solubility limits and distribution coefficients together were ob
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