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Journal Articles

Characterization of the insoluble sludge from the dissolution of irradiated fast breeder reactor fuel

Aihara, Haruka; Arai, Yoichi; Shibata, Atsuhiro; Nomura, Kazunori; Takeuchi, Masayuki

Procedia Chemistry, 21, p.279 - 284, 2016/12


 Times Cited Count:4 Percentile:92.59

Journal Articles

Flow-sheet study of MA recovery by extraction chromatography for SmART cycle project

Watanabe, So; Nomura, Kazunori; Kitawaki, Shinichi; Shibata, Atsuhiro; Kofuji, Hirohide; Sano, Yuichi; Takeuchi, Masayuki

Procedia Chemistry, 21, p.101 - 108, 2016/12


 Times Cited Count:11 Percentile:98.79

Journal Articles

Simulation study of sludge precipitation in spent fuel reprocessing

Takeuchi, Masayuki; Aihara, Haruka; Nakahara, Masaumi; Tanaka, Kotaro*

Procedia Chemistry, 21, p.182 - 189, 2016/12


 Times Cited Count:1 Percentile:67.05

A simulation technology with electrolyte thermodynamic model has been developed to evaluate the precipitation behavior in reprocessing solution based on nitric acid solution. The simulation results were compared with the experiment data from non-radioactive simulated HLLW with ten elements and Pu-Zr-Mo solution, and the reliability of the thermodynamic model was verified. Most of the precipitation species was zirconium molybdate hydrate from the both data. It is demonstrated that the chemical species and amount of the precipitation calculated by thermodynamic model reflected well that of experiments. This study has shown the thermodynamic simulation model is one of the useful tools to estimate the behavior of precipitation from the reprocessing solution.

Journal Articles

Effect of sludge behavior on performance of centrifugal contactor

Sakamoto, Atsushi; Sano, Yuichi; Takeuchi, Masayuki

Procedia Chemistry, 21, p.495 - 502, 2016/12


 Times Cited Count:2 Percentile:82.38

Journal Articles

Uranium and plutonium extraction by ${it N,N}$-dialkylamides using multistage mixer-settler extractors

Ban, Yasutoshi; Hotoku, Shinobu; Tsutsui, Nao; Suzuki, Asuka; Tsubata, Yasuhiro; Matsumura, Tatsuro

Procedia Chemistry, 21, p.156 - 161, 2016/12

 Times Cited Count:5 Percentile:94.78

A continuous counter-current experiment was carried out to demonstrate the validity of a process using ${it N,N}$-dialkylamides for recovering U and Pu. This process consisted of two cycles, and the 1st cycle and the 2nd cycle employed ${it N,N}$-di(2-ethylhexyl)-2,2-dimethylpropanamide and ${it N,N}$-di(2-ethylhexyl)butanamide as extractants, respectively. The feed solution for the 1st cycle was 5.1 mol/dm$$^{3}$$ (M) nitric acid containing 0.92 M U, 1.6 mM Pu, and 0.6 mM Np. The raffinate collected in the 1st cycle was used as the feed for the 2nd cycle. The ratios of U recovered in the U fraction and U-Pu fraction were 99.1% and 0.8%, respectively. The ratio of Pu recovered in the U-Pu fraction was 99.7%. The concentration ratio of U with respect to Pu in the U-Pu fraction was 9, and this indicated that Pu was not isolated. The decontamination factor of U with respect to Pu in the U fraction was obtained as 4.5$$times$$10$$^{5}$$. These results supported the validity of the proposed process.

Journal Articles

Characteristics of dicesium plutonium(IV) nitrate formation in separation system of uranyl nitrate hexahydrate crystal

Nakahara, Masaumi; Yano, Kimihiko; Shibata, Atsuhiro; Takeuchi, Masayuki; Okano, Masanori; Kuno, Takehiko

Procedia Chemistry, 7, p.282 - 287, 2012/00

 Times Cited Count:1 Percentile:57.41

For decontamination of Cs and Pu compound, Cs$$_{2}$$Pu(NO$$_{3}$$)$$_{6}$$, precipitated in the U cooling crystallization method, solubility measurement of Cs$$_{2}$$Pu(NO$$_{3}$$)$$_{6}$$ in a uranyl nitrate solution and a U crystallization experiments were carried out with the dissolver solution derived from irradiated fast neutron reactor core fuel. The solubility of Cs$$_{2}$$Pu(NO$$_{3}$$)$$_{6}$$ in the uranyl nitrate solution decreased with decreasing temperature. In the crystallization experiments, the decontamination factors of Cs and Pu for uranyl nitrate hexahydrate crystal decrease with increasing the Cs concentration in the feed solution because Cs$$_{2}$$Pu(NO$$_{3}$$)$$_{6}$$ formed in the course of U crystallization. Basic data were obtained for the formation behavior of Cs$$_{2}$$Pu(NO$$_{3}$$)$$_{6}$$ in the U crystallization process.

Journal Articles

Dissolution behavior of irradiated mixed-oxide fuels with different plutonium contents

Ikeuchi, Hirotomo; Shibata, Atsuhiro; Sano, Yuichi; Koizumi, Tsutomu

Procedia Chemistry, 7, p.77 - 83, 2012/00

 Times Cited Count:18 Percentile:97.37

The effects of Pu content were studied on the dissolution rate of irradiated mixed oxide fuel and on the mass of insoluble residue. Kinetic analysis was conducted being based on the surface-reaction model to estimate the dissolution rate of irradiated fuels with Pu contents less than 30% and with burn-up ranging from 40.1 - 63.7 GWD/t. The dissolution rate of irradiated mixed-oxide fuels was found to decrease exponentially with an increase of the Pu content, but those were estimated to be up to 1000 times larger than those of non-irradiated fuels with the same Pu content. The amount of insoluble residue was found to increase with increase of the Pu content, possibly due to the promotion of fission product formation. Up to 1.3% of initial heavy metal was remained as the residue.

Journal Articles

Decontamination of radioactive liquid waste with hexacyanoferrate(II)

Takahatake, Yoko; Watanabe, So; Shibata, Atsuhiro; Nomura, Kazunori; Koma, Yoshikazu

Procedia Chemistry, 7, p.610 - 615, 2012/00

 Times Cited Count:10 Percentile:94.32

Journal Articles

Advanced-ORIENT cycle project; Summary of phase I fundamental studies

Koyama, Shinichi; Suzuki, Tatsuya*; Ozawa, Masaki*; Kurosawa, Kiyoko*; Fujita, Reiko*; Mimura, Hitoshi*; Okada, Ken*; Morita, Yasuji; Fujii, Yasuhiko*

Procedia Chemistry, 7, p.222 - 230, 2012/00

 Times Cited Count:2 Percentile:72.13

Adv.-ORIENT cycle strategy has been proposed as a basic concept for trinitarian research on separation, transmutation and utilization of nuclides and elements based on FBR fuel cycle. Validation of principal separation method and related safety research were performed from 2006 through 2011 as Phase I program. First, more than 90% of Cs could be recovered from the actual spent fuel [IXC(I) step]. The next is the adsorption of the platinum group metals (PGM), lanthanides, Am and Cm were separated by using a tertiary pyridine-type resin (TPR) as ion exchange steps [IXC(II, III, IV) steps]. The separated PGM metals will be supplied to the electrochemical extraction [CEE step]. As experiment for safety issues, Hastelloy-B at RT and Ta at 90$$^{circ}$$C were confirmed their anti-corrosive in highly concentrated HCl media. Thermo-chemical stability for TPR was verified. Issues to be solved for next phase based on the final results of phase I program.

Journal Articles

Fundamental research on actinide materials for sustainable fuel cycles in JAEA

Arai, Yasuo

Procedia Chemistry, 7, p.425 - 430, 2012/00

 Times Cited Count:1 Percentile:57.41

The fundamental research on actinide materials has been carried out in order to contribute to the development of future nuclear fuel cycle and actinide science database. Among actinide materials, the R&D has been focused on Pu and minor actinide (MA; Np, Am, Cm) bearing compounds. The chemical forms of actinide compounds concerned include oxides, nitrides, chlorides and alloys, which are prepared, characterized and subjected to property measurements. In this paper those results on Pu and MA bearing oxides obtained in recent several years are summarized. In addition, the possible challenges of actinide materials research to the subjects of post severe accident of Fukushima Dai-ichi Nuclear Power Station are briefly discussed.

Journal Articles

Optimizing composition of TODGA/SiO$$_{2}$$-P adsorbent for extraction chromatography process

Watanabe, So; Arai, Tsuyoshi*; Ogawa, Tsuyoshi*; Takizawa, Makoto*; Sano, Kyohei*; Nomura, Kazunori; Koma, Yoshikazu

Procedia Chemistry, 7, p.411 - 417, 2012/00

 Times Cited Count:14 Percentile:96.32

Journal Articles

Multiplier effect on separation of Am and Cm with hydrophilic and lipophilic diamides

Sasaki, Yuji; Tsubata, Yasuhiro; Kitatsuji, Yoshihiro; Sugo, Yumi; Shirasu, Noriko; Morita, Yasuji

Procedia Chemistry, 7, p.380 - 386, 2012/00

 Times Cited Count:8 Percentile:92.64

Following the nuclear properties, the different disposal methods for Am, Cm and lanthanides in HLW have been investigating, e.g., Am; transmutation, Cm; interim storage and Ln; geological disposal. The mutual separation is an important task. However, these trivalent Ln and An are difficult to separate due to their very similar chemical behavior, same oxidation state and similar ionic radii. We try to use both hydrophilic and lipophilic diamides in an extraction system simultaneously in order to attain the effective mutual separation. In this work, lipophilic DOODA or DGA are used as the extractant and some hydrophilic diamides are employed as the masking agents. The results of mutual separation of Am/Cm/Ln are discussed in this work.

Journal Articles

Chemical durability of iron-phosphate glass as the high level waste from pyrochemical reprocessing

Kofuji, Hirohide; Yano, Tetsuji*; Myochin, Munetaka; Matsuyama, Kanae*; Okita, Takeshi*; Miyamoto, Shinya*

Procedia Chemistry, 7, p.764 - 771, 2012/00

 Times Cited Count:12 Percentile:95.69

As a part of the research and development for the nuclear waste disposal concept suitable to the advanced fuel cycle systems and its performance evaluation, the iron-phosphate glass is examined as an alternative waste form for high level waste generated from pyrochemical reprocessing. In order to enhance the waste element content in the glass matrix and improve the durability of the waste form, optimization experiments of glass composition were carried out and the effect of additional other transition metal oxides was found out in this study.

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