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JAEA Reports

Data on migration analysis for TRU waste disposal; Results of plutonium solubility experiments in porewater of cement hydrates

Suguro, Toshiyasu; Nishikawa, Yoshiaki*; Watahiki, Takashi*; Kagawa, Akio

JAEA-Technology 2013-023, 22 Pages, 2013/10

JAEA-Technology-2013-023.pdf:2.41MB

For safety assessment of TRU waste disposal, solubility of plutonium was investigated under hardened cement paste porewater condition. Polycarboxylic acid compound, which have the possibility to be used for the TRU waste disposal, was selected as the cement admixture for the experiment. Initial concentration of Pu was 10$$^{-6}$$ M in the experiment. The porewater of hardened cement paste was obtained by squeezing out the kneading of ordinary portland cement and deionized water with the cement admixture. The porewater of hardened cement paste without cement admixture is also used for the experiment. The maximum experimental period was 154 days. The experiment was carried out at room temperature (298 $$pm$$ 5 K) under argon atmosphere, in which oxygen concentration was lower than 1 ppm. Pu concentration in the porewater of hardened cement paste with or without the cement admixture were in the order of 10$$^{-10}$$ mol/dm$$^{3}$$ after 154 days. This value is comparable to the solubility of Pu(IV) under high pH condition, suggesting that the solubility of Pu was not affected by the cement admixture in hardened cement paste.

JAEA Reports

Data on migration analysis for TRU waste repository; Results of the solubility experiments in the presence of organic materials used for cement additives

Suguro, Toshiyasu; Nishikawa, Yoshiaki*; Watahiki, Takashi*; Kagawa, Akio; Iijima, Kazuki

JAEA-Technology 2010-048, 32 Pages, 2011/03

JAEA-Technology-2010-048.pdf:0.89MB

Cementitious materials are considered to be necessary for the construction of TRU waste repository. The cement additives are used for cements and concretes in order to provide their fluidity. Many kinds of cement additives contain organic compounds which may increase radionuclide solubility by complex formation. Therefore, it is important to obtain the solubility data with cement additives for safety assessment of TRU waste disposal. In this work, two types of cement additives, such as sodium formaldehyde acid polymer and poly carboxylic acid polymer which are expected to be applied to the TRU waste disposal system, are selected. Since the chemical condition of the repository is considered to be reducing, the authors carried out batch-type experiments of plutonium solubility under reducing (Na$$_{2}$$S$$_{2}$$O$$_{4}$$ added as reducant) and anoxic condition ([O$$_{2}$$] $$leqq$$ 1 ppm). Other experimental conditions are (1)initial plutonium concentration; 10$$^{-6}$$ M, (2) temperature; 298$$pm$$5 K, (3)experimental period; 7, 14, 28 and 56 days and (4) molecular weight of cement additives; without fractionation $$<$$ 5,000 and $$>$$ 5,000. The plutonium concentration in the absence of the cement additives was in the order of 10$$^{-10}$$ mol dm$$^{-3}$$, while, those in the presence of cement additives were two or three orders of magnitude higher. Additionally, low molecular weight fraction of cement additives brought relatively higher plutonium concentration than high molecular weight fraction.

JAEA Reports

Data on plutonium sorption onto rock; Results of the experiment for data on plutonium sorption onto tuff under conditions of reducing and of presence of nitrate

Suguro, Toshiyasu; Nishikawa, Yoshiaki*; Komuro, Takashi*; Kagawa, Akio; Kashiwazaki, Hiroshi; Yamada, Kazuo

JAEA-Technology 2007-058, 20 Pages, 2007/11

JAEA-Technology-2007-058.pdf:3.26MB

For safety assessment of TRU waste disposal, data on sorption data of plutonium on Tuff have been obtained by a static batch-type experiment. Because the repository condition will be reducing and be affected by considerable amount of nitrate in waste, the authors carried out the experiments using Tuff under the reducing (Na$$_{2}$$S$$_{2}$$O$$_{4}$$ as added as reductant) and anoxic condition (O$$_{2}$$$$leq$$1 ppm) and solution of 0 to 0.5 M NaNO$$_{3}$$. The experimental results suggest that distribution coefficient (Kd) ranges 0.2 to 0.7 m$$^{3}$$ kg$$^{-1}$$ in case of L/S=0.1 m$$^{3}$$ kg$$^{-1}$$. Similarly the Kd ranges, 1 to 7 m$$^{3}$$ kg$$^{-1}$$ at L/S=1 m$$^{3}$$ kg$$^{-1}$$. However, almost samples of the solution after experiments were plutonium solubility less than detection limit(10$$^{-13}$$mol/dm$$^{3}$$) of alpha spectrometer. The reason, it is guessed plutonium coprecipitation with calcium hydroxide, because experiments using saturated calcium hydroxide in the liquid.

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.

JAEA Reports

None

Kagawa, Akio; Suguro, Toshiyasu; Fukumoto, Masahiro; Miyamoto, Yoichi; Nakahishi, Yoshio

PNC TN8410 95-202, 108 Pages, 1995/06

PNC-TN8410-95-202.pdf:1.9MB

None

JAEA Reports

None

Kagawa, Akio; Suguro, Toshiyasu; Fukumoto, Masahiro; Miyamoto, Yoichi; Nakahishi, Yoshio

PNC TN8410 94-281, 60 Pages, 1994/08

PNC-TN8410-94-281.pdf:1.02MB

None

Oral presentation

Solubility of actinide ions in cement pore water, 1; Effect of super plasticizer on solubility of plutonium

Suguro, Toshiyasu; Kagawa, Akio; Nishikawa, Yoshiaki*; Watahiki, Takashi*; Mihara, Morihiro; Iijima, Kazuki

no journal, , 

no abstracts in English

Oral presentation

Renovation of building L

Nakagawa, Takuya; Yamamoto, Keisuke; Suguro, Toshiyasu; Sone, Tomoyuki

no journal, , 

no abstracts in English

9 (Records 1-9 displayed on this page)
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