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JAEA Reports

An Estimation of the Release of Radioactive Iodine from the Waste Package Solidified by the Copper Matrix

Honda, Akira; Toshiyasu, Suguro,; Sasaki, Ryoichi

JNC TN8400 2004-029, 43 Pages, 2005/03

JNC-TN8400-2004-029.pdf:4.69MB

Radioactive Iodine in the spent nuclear fuel is trapped by Iodine-adsorber in the off-gas process of reprocessing plant. The radioactive iodine includes very long half-lived nuclide (I-129;Half life=1.57x10$$^{7}$$y).The I-129 cannot be expected to decay due to containment by the barrier system because of its long half life. The Iodine have soluble and poorly sorbing nature in the geological disposal condition, because the element can take the chemical form of I$$^{-}$$in the reducing condition such as the condition of deep underground. Therefore Iodine can migrate in barrier system easily and strongly contribute to the peak of dose in the performance assessment of TRU waste disposal. An effective measure for reducing the dose peak is the controlled release of Iodine from the waste package in the low flux. The solidification by the copper matrix was proposed as one of the previous controlled release technology by JNC. The release rate of I-129 from the waste package solidified by the copper matrix was estimated. The corrosion rate of copper matrix was estimated as the sum of those both in the oxdizing and reducing conditions. The rates and periods of I-129 release were estimated under the assumption of congruent release of I-129 with corrosion of the copper matrix. The total release rate and period in the FRHP groundwater case were 3.11X10$$^{6}$$Bq y$$^{-1}$$and 1.64x10$$^{7}$$y(Initiated at 10$$^{3}$$y and finalized at 1.64x10$$^{7}$$y) respectively. The total release rate and period in the SRHP groundwater case were 9.03x10$$^{8}$$Bq y$$^{-1}$$and 5.66x10$$^{4}$$y(Initiated at 10$$^{3}$$y and finalized at 5.76x10$$^{4}$$y).

JAEA Reports

Data on Plutonium Sorption onto Cementitious Materials under Conditions of Reducing and of Presence of Nitrate

Toshiyasu, Suguro,; Notoya, Shin; Nishikawa, Yoshiaki*; Nakamura, Ryosuke*; Shibutani, Tomoki; Kuroha, Mitsuhiko; Kamei, Gento

JNC TN8430 2004-004, 27 Pages, 2005/01

JNC-TN8430-2004-004.pdf:1.03MB

In terms of safety assessment of TRU waste disposal, data on plutonium sorption on cementitious materials have been obtained by means of a static batch-type experiment. Because the repository condition will be reducing and be affected by considerable amount of nitrate, the authors carried out the experiments using ordinary portland cement (OPC) under the reducing (Na$$_{2}$$S$$_{2}$$O$$_{4}$$ as added as reductant) and anoxic condition (O$$_{2}$$ $$leq$$ 1ppm) and solution of 0 to 0.5 M NaNO$$_{3}$$. Other experimental conditions are : liquid/solid (L/S) ratios ; 100 and 1000 mL g$$^{-1}$$, Initially aaded plutonium; 2.84$$times$$10$$^{-10}$$M, Temperature; 25$$pm$$5$$^{circ}$$C and Reaction times; 7, 14 and 28 days. The experimental results suggest that distribution coefficient ($$Kd$$) ranges 50 to 1000 mL g$$^{-1}$$ in case of L/S=100mL g$$^{-1}$$. Similarly the $$Kd$$ ranges, 100 to 10000 mL g$$^{-1}$$ at L/S=1000mL g$$^{-1}$$. These $$Kd$$ values tend to increase with lapsing reaction time. On the basis of these results, we recommend 50mL g$$^{-1}$$ as a conservative $$Kd$$ value of plutonium on OPC in a TRU waste repository condition.

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