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The sorption of thorium and americium has been measured onto crushed samples of freshordinary Portland cement (OPC), degraded OPC (DOPC) and green tuff under a range of aqueous conditions as part of research into the disposal of TRU waste in Japan. Sorption onto OPC was measured from deionised water,3 mol dm−3 sodium nitrate solution and simulated seawater, and onto DOPC from demineralised water; all solutions were pre-equilibrated with the relevant cement. Sorption onto tuff was measured from pre-equilibrated deionised water, 3 mol dm−3 sodium nitrate solution, simulated seawater, OPC leachate and 1 mol dm−3 ammonium hydroxide solution. RD values were determined from the amount of thorium or americium remaining in solution after centrifugation, 0.45 μm filtration and 10,000 nominal molecular weight cut-off (NMWCO) filtration. Centrifugation gave lower RD values than filtration. MeanRD values for sorption onto OPC and DOPC were in the range 4 × 104 to ≥1 × 106 cm3g−1 for thorium, and 3 ×103 to ≥1 × 105 cm3g−1 for americium, after filtration through 10,000 NMWCO filters. Mean RD values for thorium sorption onto tuff increased from2 × 103 cm3g−1 in tuff-equilibrated deionised water, to ≥ 4 × 106 cm3g−1 from ammonium hydroxide solution (10,000 NMWCO-filtration). A similar trend was seen for10,000 NMWCO-iltered americium samples where mean RD values increased from 4 × 103 cm3g−1 in deionised water to 1 × 105 cm3g−1 in ammonium hydroxide solution.
United Kingdom Nirex Limited develops and advises on safe, environmentally sound and publicly acceptable options for the long-term management of radioactive waste. One option Nirex has developed is a phased geological repository concept for intermediate level waste and some low level wastes that makes use of a combination of engineered and natural barriers. Physical containment of radionuclides would be achieved by immobilisation and packaging of wastes (mostly) in stainless steel containers.
Existing models of the migration of dissolved radionuclides from packaged wastes suggest that radionuclide release is determined largely by the rate of diffusion through the encapsulation grout used to immobilise the waste. The use of such models requires diffusion coefficient data for radionuclides in waste encapsulation grouts. This paper describes a programme of through-diffusion experiments, and modelling interpretation, aimed at deriving diffusion coefficients for some radionuclides in two types of encapsulation grout.
An intrinsic diffusion coefficient of HTO of around 1×10−13 to 2×10−13 m2s−1 was determined for a 3:1 mix of blast furnace slag to ordinary Portland cement, compared to around 4×10−13 to 5×10−13 m2s−1 for a 3:1 mix of pulverised fuel ash to ordinary Portland cement. These values are lower than that assumed for a non-sorbing radionuclide in an earlier modelling exercise. Porosity values around 0.3 were obtained in each case. For 36Cl as chloride, the experiments showed no significant breakthrough over the experimental timescale of about one year, suggesting an intrinsic diffusion coefficient below 5×10−13 m2s−1. One possibility is that chlorine-containing solids are precipitating in the cement. An intrinsic diffusion coefficient for 137Cs in the 3:1 mix of pulverised fuel ash to ordinary Portland cement of 4×10−15 m2s−1 was estimated, significantly lower than that determined for HTO.
The results from three of the sixteen experiments could not be fitted with a simple diffusion model, and for a further five experiments there was some doubt as to whether simple diffusion behaviour had been observed. It is suggested that this may have been due to cracks in the grouts that were sufficiently large to affect the diffusion properties of the grouts, although none was visible to the naked eye. Cracking of the waste encapsulation grouts could provide a mechanism for enhanced migration of radionuclides from waste packages, compared with diffusion in a homogeneous porous medium alone.
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