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Kofuji, Hirohide; Yano, Tetsuji*; Myochin, Munetaka; Matsuyama, Kanae*; Okita, Takeshi*; Miyamoto, Shinya*
Procedia Chemistry, 7, p.764 - 771, 2012/00
Times Cited Count：14 Percentile：95.27(Chemistry, Analytical)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.
Arai, Yasuo
Procedia Chemistry, 7, p.425 - 430, 2012/00
Times Cited Count：1 Percentile：55.31(Chemistry, Analytical)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.
-P adsorbent for extraction chromatography processWatanabe, So; Arai, Tsuyoshi*; Ogawa, Tsuyoshi*; Takizawa, Makoto*; Sano, Kyohei*; Nomura, Kazunori; Koma, Yoshikazu
Procedia Chemistry, 7, p.411 - 417, 2012/00
Times Cited Count：15 Percentile：95.90(Chemistry, Analytical)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：91.69(Chemistry, Analytical)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.
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：55.31(Chemistry, Analytical)For decontamination of Cs and Pu compound, CsPu(NO
)
, precipitated in the U cooling crystallization method, solubility measurement of CsPu(NO) 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 CsPu(NO) 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 CsPu(NO) formed in the course of U crystallization. Basic data were obtained for the formation behavior of CsPu(NO) in the U crystallization process.
Ikeuchi, Hirotomo; Shibata, Atsuhiro; Sano, Yuichi; Koizumi, Tsutomu
Procedia Chemistry, 7, p.77 - 83, 2012/00
Times Cited Count：20 Percentile：97.37(Chemistry, Analytical)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.
Takahatake, Yoko; Watanabe, So; Shibata, Atsuhiro; Nomura, Kazunori; Koma, Yoshikazu
Procedia Chemistry, 7, p.610 - 615, 2012/00
Times Cited Count：12 Percentile：94.64(Chemistry, Analytical)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：70.03(Chemistry, Analytical)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
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.
Kofuji, Hirohide; Yano, Tetsuji*; Myochin, Munetaka; Okita, Takeshi*; Miyamoto, Shinya*
no journal, ,
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 in this study.
Kitawaki, Shinichi; Nakayoshi, Akira; Fukushima, Mineo; Murakami, Tsuyoshi*; Sakamura, Yoshiharu*; Akiyama, Naoyuki*
no journal, ,
no abstracts in English
Murakami, Tsuyoshi*; Sakamura, Yoshiharu*; Akiyama, Naoyuki*; Kitawaki, Shinichi; Nakayoshi, Akira; Koyama, Tadafumi*
no journal, ,
no abstracts in English
Onishi, Takashi; Koyama, Shinichi; Mimura, Hitoshi*
no journal, ,
Fission products are generated by fission reaction in a nuclear fuel. Platinum group elements, such as ruthenium (Ru), rhodium (Rh) and palladium (Rh) are also produced. Their elements play important roles in chemical and electrical industry. Mimura et al. developed hybrid microcapsules for recovery of these useful platinum group elements from high-level liquid waste (HLLWs). The hybrid microcapsules enclosing KCuFC (potassium copper hexacyanoferrate) immobilized by alginate gel were prepared. The hybrid microcapsules adsorbed Ru, Rh and Pd in single element solutions. Recovery experiments from simulated high level radioactive liquid waste were conducted in this study. Most of the Ru, Rh and Pd were adsorbed together on the hybrid microcapsules from the simulated HLLWs. A conversion method of Pd solid from solution was also proposed for effective use of recovered palladium. The separated Pd solution was converted to solid materials by thermal decomposition and acid digestion.
